Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Shinohara, Masanori; Sumita, Junya; Shibata, Taiju; Hirata, Masaru
JAEA-Evaluation 2022-006, 198 Pages, 2022/11
JAEA-Evaluation-2022-006.pdf:23.34MB
Japan Atomic Energy Agency (JAEA) received a post-evaluation of the third medium-/long-term plan (from FY2015 to FY2021) and pre-evaluation of the fourth medium-/long-term plan (from FY2022 to FY2028) from the "Evaluation Committee of Research Activities for High Temperature Gas-cooled Reactor (hereinafter referred to as "HTGR") and Related Hydrogen Production Technology" (hereinafter referred to as "Evaluation Committee") which consists of specialists in the fields of evaluation subjects of HTGR and related heat application technologies. As a result, for the post-evaluation of the third medium-/ long-term plan, two of the ten technical committee members concluded a score of "S", seven members concluded "A" and one member concluded "B". The comprehensive evaluation concluded a score of "A". On the other hand, one of the two humanities and social sciences members concluded a score of "B", one members concluded "C". The comprehensive evaluation concluded a score of "B". For the pre-evaluation of the fourth medium-/long-term plan, although there were some items that several evaluation committee members rated as "needs improvement," the majority of the committee members judged the plan to be appropriate. This report describes the members of the Evaluation Committee, assessment items, assessment results and JAEA's measures following the assessment.
Shinohara, Masanori; Sumita, Junya; Inaba, Yoshitomo; Shibata, Taiju
Dai-59-Kai X Sen Zairyo Kyodo Ni Kansuru Toronkai Koen Rombunshu, p.22 - 28, 2022/11
no abstracts in English
Matsuba, Kenichi; Shinohara, Masanori; Toyooka, Junichi; Inaba, Yoshitomo; Sumita, Junya
Enerugi, Shigen, 43(4), p.218 - 223, 2022/07
In the global trend toward decarbonization, Japan has a policy to pursue all options, including nuclear power, to achieve carbon neutrality by 2050. In order to meet the public requirements for nuclear power, it is important to promote the development of advanced reactors, including the Small Modular Reactor (SMR), as one of the promising options. This article describes the domestic and international trends of SMR development, introduces the activities of Japan Atomic Energy Agency (JAEA) for the development of advanced reactors including SMRs, and concludes with the future prospect for the introduction of advanced reactor including SMRs in Japan.
Shinohara, Masanori; Sumita, Junya; Shibata, Taiju; Hirata, Masaru
JAEA-Evaluation 2022-001, 104 Pages, 2022/06
JAEA-Evaluation-2022-001.pdf:28.15MB
Japan Atomic Energy Agency received the annual evaluation of FY2020, the research plan of FY2021 and the prospective evaluation of the third mid- to long-term plan (from FY2015 to FY2021) from the "Evaluation Committee of Research Activities for High Temperature Gas-cooled Reactor (hereinafter referred to as "HTGR") and Related Hydrogen Production Technology" (hereinafter referred to as "Evaluation Committee") which consists of specialists in the fields of evaluation subjects of HTGR and related heat application technology. As a result, for the annual evaluation of FY2020, one of the ten technical committee members concluded a score of "S", eight members concluded "A" and one member concluded "B". The comprehensive evaluation was concluded a score of "A". On the other hand, two humanities and social sciences committee members concluded "B". For the prospective evaluation of the third mid- to long-term plan, one of the ten technical committee members concluded a score of "S", eight members concluded "A" and one member concluded "B". The comprehensive evaluation was concluded a score of "A". On the other hand, two humanities and social sciences committee members concluded "B". This report describes the members of the Evaluation Committee, assessment items, assessment results and JAEA's measures following the assessment.
Yonomoto, Taisuke; Nakashima, Hiroshi*; Sono, Hiroki; Kishimoto, Katsumi; Izawa, Kazuhiko; Kinase, Masami; Osa, Akihiko; Ogawa, Kazuhiko; Horiguchi, Hironori; Inoi, Hiroyuki; et al.
JAEA-Review 2020-056, 51 Pages, 2021/03
JAEA-Review-2020-056.pdf:3.26MB
A group named as "The group for investigation of reasonable safety assurance based on graded approach", which consists of about 10 staffs from Sector of Nuclear Science Research, Safety and Nuclear Security Administration Department, departments for management of nuclear facility, Sector of Nuclear Safety Research and Emergency Preparedness, aims to realize effective graded approach (GA) about management of facilities and regulatory compliance of JAEA. The group started its activities in September, 2019 and has had discussions through 10 meetings and email communications. In the meetings, basic ideas of GA, status of compliance with new regulatory standards at each facility, new inspection system, etc were discussed, while individual investigation at each facility were shared among the members. This report is compiled with expectation that it will help promote rational and effective safety management based on GA by sharing contents of the activity widely inside and outside JAEA.
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 1
C. 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.
Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji
JAEA-Technology 2016-038, 36 Pages, 2017/02
JAEA-Technology-2016-038.pdf:8.75MB
The High Temperature Engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of 
Cf with 3.7GBq The NSs are exchanged at the interval of approximately 7 years. The NS holders including NSs are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the Control Rod guide blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Following technical issues were extracted from the experiences in the past two exchange works of NSs to develop a safety handling procedure; (1) The reduction and prevention of radiation exposure of workers. (2) The exclusion of falling of NS holder. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container and satisfying the regulation of A type transportation package.
Shimazaki, Yosuke; Sawahata, Hiroaki; Shinohara, Masanori; Yanagida, Yoshinori; Kawamoto, Taiki; Takada, Shoji
Journal of Nuclear Science and Technology, 54(2), p.260 - 266, 2017/02
Times Cited Count：2 Percentile：19.65(Nuclear Science & Technology)The High-Temperature engineering Test Reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of Cf with 3.7 GBq and is contained in a small capsule, installed in NS holder and subsequently in a control guide block (CR block). The NSs are exchanged at the interval of approximately 7 years. The NS holders are transported from the dealer's hot cell to the reactor facility of HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR. Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container. As the results, the NS handling work using the new transportation container was safely accomplished by developing the new transportation container which can reduce the risks of radiation exposure dose of workers and exclude the falling of NS holder.
Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
JAEA-Technology-2016-033.pdf:11.14MB
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
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.
Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 2(4), p.041008_1 - 041008_5, 2016/10
Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.
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.
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.
Shimazaki, Yosuke; Sawahata, Hiroaki; Kawamoto, Taiki; Suzuki, Hisashi; Shinohara, Masanori; Honda, Yuki; Katsuyama, Kozo; Takada, Shoji; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
Maintenance technologies for the reactor system have been developed by using the high-temperature engineering test reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for the future HTGRs by shifting from the time-based maintenance to condition-based maintenance. The technical issue of the maintenance of in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because that it is difficult to observe directly inside of the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside of the reactor by using the time domain reflectometry and so on. The disconnection position, which was specified by the electrical method, was identified by non-destructive and destructive inspection. The accumulated data is expected to be contributed for advanced maintenance of future HTGRs.
Takada, Shoji; Shimizu, Atsushi; Kondo, Makoto; Shimazaki, Yosuke; Shinohara, Masanori; Seki, Tomokazu; Tochio, Daisuke; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
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 demonstrate the inherent safety of HTGR be secured by natural phenomena to make it possible to design a severe accident free reactor. 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. 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.
Shinohara, Masanori; Inaba, Yoshitomo; Hamamoto, Shimpei; Fujimoto, Nozomu
Journal of Nuclear Science and Technology, 51(11-12), p.1398 - 1406, 2014/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In the primary cooling system of the High Temperature engineering Test Reactor (HTTR) with an outlet coolant temperature of 950C, 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.
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：53.31(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.
Ono, Masato; Shinohara, Masanori; Iigaki, Kazuhiko; Tochio, Daisuke; Nakagawa, Shigeaki; Shimazaki, Yosuke
JAEA-Technology 2013-042, 45 Pages, 2014/01
JAEA-Technology-2013-042.pdf:10.51MB
In HTTR, it has passed about two years since the last performance confirmation test. During two years, the integrity of active equipment, leakage efficiency of coolant pressure boundary of piping and vessel and control system performance due to influence of damage and deterioration by earthquake and aging were not confirmed. To confirm them, the cold test by using HTTR was conducted and the system performances such as above mentioned items were evaluated by comparing with the plant data obtained by the past cold test. In the result, no abnormity was found in all the data in the cooling system of HTTR, and it was confirmed that the integrity of facilities and instruments of HTTR was maintained in good condition.
