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Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(12), p.820 - 824, 2021/12
The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.
Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.
High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02
As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950
C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.
Ono, Masato; Iigaki, Kazuhiko; Sawahata, Hiroaki; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Kondo, Toshinari; Kojima, Keidai; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 4(2), p.020906_1 - 020906_8, 2018/04
On March 11th, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the High Temperature Engineering Test Reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the "visual inspection of equipment and instruments". The other is the "seismic response analysis" for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.
Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Takada, Shoji
Journal of Nuclear Engineering and Radiation Science, 3(4), p.041013_1 - 041013_8, 2017/10
In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.
Ono, Masato; Fujiwara, Yusuke; Honda, Yuki; Sato, Hiroyuki; Shimazaki, Yosuke; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04
Japan Atomic Energy Agency (JAEA) has carried out research and developments towards nuclear heat utilization of High Temperature Gas-cooled Reactor (HTGR) using High Temperature Engineering Test Reactor (HTTR). The nuclear heat utilization systems connected to HTGR will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations continue even if abnormal events occur in the systems. Heat application system abnormal simulating test with HTTR was carried out in non-nuclear heating operation to focus on the thermal effect in order to obtain data of the transient temperature behavior of the metallic components in the Intermediate Heat Exchanger (IHX). The IHX is the key components to connect the HTTR with the heat application system. In the test, the coolant helium gas temperature was heated up to 120C by the compression heat of the gas circulators in the HTTR under the ideal condition to focus on the heat transfer. The tests were conducted by decreasing the helium gas temperature stepwise by increasing the mass flow rate to the air cooler. The temperature responses of the IHX were investigated. For the components such as the heat transfer tubes and heat transfer enhancement plates of IHX, the temperature response was slower in the lower position in comparison with the higher position. The reason is considered that thermal load fluctuation is imposed in the secondary helium gas which flows from the top to the bottom in the heat transfer tubes of the IHX. The test data are useful to verify the numerical model of the safety evaluation code.
Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki
JAEA-Technology 2016-025, 42 Pages, 2016/11
JAEA-Technology-2016-025.pdf:17.88MB
A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.
Ono, Masato; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 2(4), p.044502_1 - 044502_4, 2016/10
In the loss of forced core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of reactor core is stopped without inserting control rods into the core and cooling by Vessel Cooling System (VCS) to verify safety evaluation codes to investigate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water cooling tube by thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from gas circulators with stopping water flow in the VCS, the local higher temperature position was specified although the temperature was sufficiently lower than the maximum allowable working temperature, and natural circulation of water had insufficient cooling effect on the temperature of water cooling tube below 1C. Then, a new safe and secured procedure for the loss of forced core cooling test was established, which will be carried out soon after the restart of HTTR.
Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Shimizu, Atsushi; Inoi, Hiroyuki; Tochio, Daisuke; Hamamoto, Shimpei; Nishihara, Tetsuo; Takada, Shoji; Sawa, Kazuhiro; et al.
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 12 Pages, 2016/06
On March 11th, 2011, the Great East Japan Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the HTTR had been stopped under the periodic inspection and maintenance of equipment and instrument. In the great earthquake, the maximum seismic acceleration observed at the HTTR exceeded the maximum value in seismic design. The visual inspection of HTTR facility was carried out for the seismic integrity conformation of HTTR. The seismic analysis was also carried out using the observed earthquake motion at HTTR site to confirm the integrity of HTTR. The concept of comprehensive integrity evaluation for the HTTR facility is divided into two parts. One is the inspection of equipment and instrument. The other is the seismic response analysis using the observed earthquake. For the basic inspections of equipment and instrument were performed for all them related to the operation of reactor. The integrity of the facilities is confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the result of inspection of equipment and instrument and seismic response analysis, it was judged that there was no problem to operate the reactor, because there was no damage and performance deterioration, which affects the reactor operation. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013 and 2015.
Fujiwara, Yusuke; Nemoto, Takahiro; Tochio, Daisuke; Shinohara, Masanori; Ono, Masato; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
In HTTR, the test was carried out at the reactor thermal power of 9 MW under the condition that one cooling line of VCS was stopped to simulate the partial loss of cooling function from the surface of RPV in addition to the loss of forced cooling flow in the core simulation. The test results showed that temperature change of the core internal structures and the biological shielding concrete was slow during the test. Temperature of RPV decreased several degrees during the test. The temperature decrease of biological shielding made of concrete was within 1C. The numerical result simulating the detail configuration of the cooling tubes of VCS showed that the temperature rise of cooling tubes of VCS was about 15 degree C, which is sufficiently small, which did not significantly affect the temperature of biological shielding concrete. As the results, it was confirmed that the cooling ability of VCS can be kept in case that one cooling line of VCS is lost.
Tochio, Daisuke; Fujiwara, Yusuke; Ono, Masato; Shinohara, Masanori; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 9 Pages, 2016/06
From the HTTR operational experience, it is needed to clear the thermal mixing characteristics of the helium gas at the annulus of the co-axial double-walled piping in HTGR. In this paper, thermal-hydraulic analysis on the helium gas at the annular flow path of the co-axial double pipe with T-junction was conducted. The analysis was performed under the condition of the different annular flow path height and with the different flow rate of the higher- and the lower-temperature helium gas. It is shown that the thermal mixing behavior is not so much affected by the flow rate of higher- and lower-temperature helium gas, and it is difficult to mix the helium gas with the smaller height of the annular flow path. It is confirmed that it is difficult to mix the helium gas in the annular flow path of the co-axial double-walled piping by using the hydraulic behavior, and it is necessary to arrange the mixing promotor in the annular flow path.
Honda, Yuki; Tochio, Daisuke; Sato, Hiroyuki; Nakagawa, Shigeaki; Ono, Masato; Fujiwara, Yusuke; Hamamoto, Shimpei; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 5 Pages, 2016/06
The characteristic confirmation test has been demonstrating by using the High Temperature engineering Test Reactor (HTTR). The thermal load fluctuation test, which is one of marginal performance test is planned to be carried out after restarting of the HTTR. The preliminary analysis for the thermal load fluctuation test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by thermal fluctuation. It means that HTGR have the capability of absorbing thermal fluctuation. This paper focuses on the investigation of mechanism of absorbing thermal fluctuation. With additional analysis, it is cleared that the large negative graphite moderator reactivity enhances the capability of absorbing thermal fluctuation. In addition, in the middle of the core, graphite moderator reactivity insertion trend are inverted. This trend is unique to HTGR because of large temperature difference between core inlet and outlet.
Ono, Masato; Iigaki, Kazuhiko; Shimazaki, Yosuke; Tochio, Daisuke; Shimizu, Atsushi; Inoi, Hiroyuki; Hamamoto, Shimpei; Takada, Shoji; Sawa, Kazuhiro
Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.363 - 371, 2016/03
HTTR is graphite moderated and helium gas-cooled reactor with prismatic fuel elements and hexagonal blocks. Here, the graphite block is brittle materials and might be damaged by collision of neighboring blocks by the large earthquake. A seismic observation system is installed in the HTTR site to confirm a behavior of a seismic event. On March 11th, 2011, off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. After the accident at the TEPCO Fukushima Daiichi Nuclear Power Station, the safety of nuclear reactors is the highest importance. To confirm the seismic integrity of HTTR core components, the seismic analysis was carried out using the evaluation waves based on the relationship between the observed earthquake motion at HTTR site and frequency transfer function. In parallel, confirmation tests of primary cooling system on cold state and integrity confirmation of reactor buildings and component support structures were also carried out. As a result, it was found that a stress value of the graphite blocks satisfied an allowable value, and the integrity of the HTTR core components was ensured. The integrity of HTTR core components was also supported by the operation without reactor power in cold conditions of HTTR. The obtained data was compared with the normal plant data before the earthquake. As the result, the integrity of the HTTR facilities was confirmed.
Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Takada, Shoji; Sawahata, Hiroaki; Kawamoto, Taiki; Hamamoto, Shimpei; Shinohara, Masanori
Proceedings of International Topical Meeting on Research Reactor Fuel Management and Meeting of the International Group on Reactor Research (RRFM/IGORR 2016) (Internet), p.1034 - 1042, 2016/03
In High Temperature Engineering Test Reactor (HTTR), three neutron holders containing 
Cf with 3.7 GBq for each are loaded in the graphite blocks and inserted into the reactor core as a neutron startup source which is changed at the interval of approximately ten years. These neutron holders containing the neutron sources are transported from the dealer's hot cell to HTTR using the transportation container. The holders loading to the graphite block are carried out in the fuel handling machine maintenance pit of HTTR. There were two technical issues for the safety handling work of the neutron holder. The one is the radiation exposure caused by significant movement of the container due to an earthquake, because the conventional transportation container was so large (
1240 mm, h1855 mm) that it can not be fixed on the top floor of maintenance pit by bolts. The other is the falling of the neutron holder caused by the difficult remote handling work, because the neutron holder capsule was also so long (155 mm, h1285 mm) that it can not be pulled into the adequate working space in the maintenance pit. Therefore, a new and low cost transportation container, which can solve the issues, was developed. To avoid the neutron and 
ray exposure, smaller transportation container (820mm, h1150 mm) which can be fixed on the top floor of maintenance pit by bolts was developed. In addition, to avoid the falling of the neutron holder, smaller neutron holder capsule (75 mm, h135 mm) with simple handling mechanism which can be treated easily by manipulator was also developed. As the result of development, the neutron holder handling work was safely accomplished. Moreover, a cost reduction for manufacturing was also achieved by simplifying the mechanism of neutron holder capsule and downsizing.
Nakashima, Yosuke*; Takeda, Hisahito*; Ichimura, Kazuya*; Hosoi, Katsuhiro*; Oki, Kensuke*; Sakamoto, Mizuki*; Hirata, Mafumi*; Ichimura, Makoto*; Ikezoe, Ryuya*; Imai, Tsuyoshi*; et al.
Journal of Nuclear Materials, 463, p.537 - 540, 2015/08
Times Cited Count：20 Percentile：83.66(Materials Science, Multidisciplinary)Nakashima, Yosuke*; Sakamoto, Mizuki*; Yoshikawa, Masayuki*; Oki, Kensuke*; Takeda, Hisahito*; Ichimura, Kazuya*; Hosoi, Katsuhiro*; Hirata, Mafumi*; Ichimura, Makoto*; Ikezoe, Ryuya*; et al.
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Iigaki, Kazuhiko; Ono, Masato; Shimazaki, Yosuke; Tochio, Daisuke; Shimizu, Atsushi; Inoi, Hiroyuki; Takada, Shoji; Sawa, Kazuhiro
Mechanical Engineering Journal (Internet), 1(4), p.TEP0021_1 - TEP0021_13, 2014/08
On March 11th, 2011, the Great East Japan Earthquake which is one of the largest earthquakes in Japan occurred and the maximum acceleration in observed seismic wave in the HTTR exceeded the design value in a part of input seismic motions. Therefore, a visual inspection, a seismic analysis and a performance confirmation test of facilities were carried out in order to confirm the integrity of facility after the earthquake. The seismic analysis was carried out for the reactor core structures by using the response magnification factor method. As the results of the evaluation, the generated stress in the graphite blocks in the reactor core at the earthquake were well below the allowable values of safety criteria, and thus the structural integrity of the reactor core was confirmed. The integrity of reactor core was also supported by the visual inspections of facilities and the operation without reactor power in cold conditions of HTTR.
Takada, Shoji; Iigaki, Kazuhiko; Shinohara, Masanori; Tochio, Daisuke; Shimazaki, Yosuke; Ono, Masato; Yanagi, Shunki; Nishihara, Tetsuo; Fukaya, Yuji; Goto, Minoru; et al.
Nuclear Engineering and Design, 271, p.472 - 478, 2014/05
Times Cited Count：8 Percentile：53.13(Nuclear Science & Technology)JAEA has carried out research and development to establish the technical basis of HTGRs using HTTR. To connect hydrogen production system to HTTR, it is necessary to ensure the reactor dynamics when thermal-load of the system is lost. Thermal-load fluctuation test is planned to demonstrate the reactor dynamics stability and to validate plant dynamics codes. It will be confirmed that the reactor become stable state during losing a part of removed heat at heat-sink. A temperature coefficient of reactivity is one of the important parameters for core dynamics calculations, and changes with burnup because of variance of fuel compositions. Measurement of temperature coefficient of reactivity has been conducted to confirm the validity of calculated temperature coefficient of reactivity. A LOFC test using HTTR has been carried out to verify the inherent safety under the condition of LOFC while the reactor shut-down system disabled.
Takada, Shoji; Shinohara, Masanori; Seki, Tomokazu; Shimazaki, Yosuke; Ono, Masato; Tochio, Daisuke; Iigaki, Kazuhiko; Sawa, Kazuhiro
JAEA-Technology 2014-001, 34 Pages, 2014/03
JAEA-Technology-2014-001.pdf:4.46MB
The loss of forced cooling with vessel cooling system inactive has been planned by using HTTR at the reactor power 9 MW. In this test, the forced cooling of reactor core is lost and the vessel cooling system which removes decay heat from core is tripped. In the test, the technical items such that the temperature of water cooling tubes is expected to be higher are considered. The methods to solve such technical items were proposed. The proposed methods were verified based on the test data of the cold test toward the proposal of test plan of safety demonstration test. In the cold test, the two water trains of vessel cooling system was tripped under the condition that the reactor was heated up without nuclear heating. The reactor inlet temperature was set at 120 and 150C.
Tokuhisa, Atsushi*; Arai, Junya*; Jochi, Yasumasa*; Ono, Yoshiyuki*; Kameyama, Toyohisa*; Yamamoto, Keiji*; Hatanaka, Masayuki*; Gerofi, B.*; Shimada, Akio*; Kurokawa, Motoyoshi*; et al.
Journal of Synchrotron Radiation, 20(6), p.899 - 904, 2013/11
Times Cited Count：5 Percentile：29.09(Instruments & Instrumentation)