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Takamatsu, Kuniyoshi
Annals of Nuclear Energy, 106, p.71 - 83, 2017/08
The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.
Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06
The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.
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 1
C. 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.
Kobayashi, Masato*; Saito, Masahiko*; Iwatani, Takafumi*; Nakayama, Masashi; Tanai, Kenji; Fujita, Tomoo; Asano, Hidekazu*
JAEA-Research 2015-018, 14 Pages, 2015/12
JAEA-Research-2015-018.pdf:5.43MB
JAEA and RWMC concluded the letter of cooperation agreement on the research and development of radioactive waste disposal in April, 2005, and have been carrying out the collaboration work based on the agreement. JAEA have been carrying out the Horonobe URL Project which is intended for a sedimentary rock in the Horonobe town, Hokkaido, since 2001. In the project, geoscientific research and research and development on geological disposal technology are being promoted. Meanwhile, The Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry has been promoting construction of equipments for the full-scale demonstration of engineered barrier system and operation technology for high-level radioactive waste disposal since 2008, to enhance public's understanding to the geological disposal of HLW, e.g. using underground facility. RWMC received an order of the project in fiscal year 2014 continuing since fiscal year 2008. Since topics in this project are included in the Horonobe URL Project, JAEA carried out this project as collaboration work continuing since fiscal year 2008. This report summarizes the results of the research on engineering technology carried out in this collaboration work in fiscal year 2014.
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.
Fujita, Tomoo; Tanai, Kenji; Nakayama, Masashi; Sawada, Sumiyuki*; Asano, Hidekazu*; Saito, Masahiko*; Yoshino, Osamu*; Kobayashi, Masato*
JAEA-Research 2014-031, 44 Pages, 2015/03
JAEA-Research-2014-031.pdf:16.11MB
Japan Atomic Energy Agency (JAEA) and Radioactive Waste Management Funding and Research Center (RWMC) concluded the letter of cooperation agreement on the research and development of radioactive waste disposal in April, 2005, and have been carrying out the collaboration work based on the agreement. JAEA have been carrying out the Horonobe Underground Research Laboratory (URL) Project which is intended for a sedimentary rock in the Horonobe town, Hokkaido, since 2001. In the project, geoscientific research and research and development on geological disposal technology are being promoted. Meanwhile, the government (the Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry) has been promoting construction of equipments for the full-scale demonstration of engineered barrier system (EBS) and operation technology for high-level radioactive waste (HLW) disposal since 2008, to enhance public's understanding to the geological disposal of HLW, e.g. using underground facility. RWMC received an order of the project in fiscal year 2012 (2011/2012) continuing since fiscal year 2008 (2008/2009). Since topics in this project are included in the Horonobe URL Project, JAEA carried out this project as collaboration work continuing since fiscal year 2008. This report summarizes the results of engineering technology carried out in this collaboration work in fiscal year 2013. In fiscal year 2013, emplacement tests using buffer material block for the vertical emplacement concept were carried out and visualization tests for water penetration in buffer material were carried out.
Takamatsu, Kuniyoshi; Tochio, Daisuke; Nakagawa, Shigeaki; Takada, Shoji; Yan, X.; Sawa, Kazuhiro; Sakaba, Nariaki; Kunitomi, Kazuhiko
Journal of Nuclear Science and Technology, 51(11-12), p.1427 - 1443, 2014/11
Times Cited Count：13 Percentile：70.2(Nuclear Science & Technology)In a safety demonstration test involving a loss of both reactor reactivity control and core cooling, HTGRs such as the HTTR, which is the only HTGR in Japan, demonstrate that the reactor power would stabilize spontaneously. In the test at an initial power of 30%, when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero, a reactor transient was initiated and examined. The results confirmed that the reactor power would decrease immediately and become effectively zero.
Sakamoto, Mizuki*; Ono, Noriyasu*; Asakura, Nobuyuki; Hoshino, Kazuo
Purazuma, Kaku Yugo Gakkai-Shi, 90(8), p.473 - 479, 2014/08
no abstracts in English
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03
The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. A one-point core dynamics approximation with one fuel channel model had applied to this analysis. It was found that the analytical model for core dynamics couldn't simulate the reactor power behavior accurately. This report proposes an original method using temperature coefficients of some regions in the core. It is crucial to evaluate this method precisely to simulate a performance of HTGR during the test.
Isono, Takaaki; Koizumi, Norikiyo; Okuno, Kiyoshi; Kurihara, Ryoichi; Nishio, Satoshi; Tobita, Kenji
Fusion Engineering and Design, 81(8-14), p.1257 - 1261, 2006/02
Times Cited Count：6 Percentile：36.38(Nuclear Science & Technology)In order to realize an economically competitive power generation system, generation of a higher field is required. Toroidal Field (TF) coils of fusion DEMO plant at JAERI are required to generate magnetic field of 16 to 20 T. To realize this high field, advanced superconducting materials, such as Nb
Al and high temperature superconductor (HTS), are considered. HTS has enough performance in a 20-T field at 4 K, and a forced-cooled type HTS conductor using a silver alloy sheathed Bi-2212 round wire has been proposed. Required areas of superconductor, structure, stabilizer, coolant and insulator in the cross section of coil winding have been calculated. However, there are many technical issues to be solved, such as accurate temperature control during heat treatment in an atmosphere of oxygen. On the other hand, a large coil using NbAl has been developed by JAERI, and major technology to fabricate a 16-T NbAl coil was developed. Validity and issues of grading the winding area are discussed, and there is a possibility to increase a field up to around 17 T using the method.
Sakamoto, Keishi; Takahashi, Koji; Kasugai, Atsushi; Minami, Ryutaro; Kobayashi, Noriyuki*; Nishio, Satoshi; Sato, Masayasu; Tobita, Kenji
Fusion Engineering and Design, 81(8-14), p.1263 - 1270, 2006/02
Times Cited Count：6 Percentile：36.38(Nuclear Science & Technology)no abstracts in English
Inoue, Takashi; Hanada, Masaya; Kashiwagi, Mieko; Nishio, Satoshi; Sakamoto, Keishi; Sato, Masayasu; Taniguchi, Masaki; Tobita, Kenji; Watanabe, Kazuhiro; DEMO Plant Design Team
Fusion Engineering and Design, 81(8-14), p.1291 - 1297, 2006/02
Times Cited Count：11 Percentile：60.27(Nuclear Science & Technology)Requirement and technical issues of the neutral beam inejctor (NBI) is discussed for fusion DEMO plant. The NBI for the fusion DEMO plant should be high efficiency, high energy and high reliability with long life. From the view point of high efficiency, use of conventional electrostatic accelerator is realistic. Due to operation under radiation environment, vacuum insulation is essential in the accelerator. According to the insulation design guideline, it was clarified that the beam energy of 1.5
2 MeV is possible in the accelerator. Development of filamentless, and cesium free ion source is required, based on the existing high current/high current density negative ion production technology. The gas neutralization is not applicable due to its low efficiency (60%). R&D on an advanced neutralization scheme such as plasma neutralization (efficiency: 
80%) is required. Recently, development of cw high power semiconductor laser is in progress. The paper shows a conceptual design of a high efficiency laser neutralizer utilizing the new semiconductor laser array.
Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato
Fusion Engineering and Design, 75-79, p.313 - 318, 2005/11
Times Cited Count：10 Percentile：56.74(Nuclear Science & Technology)no abstracts in English
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Iyoku, Tatsuo
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11) (CD-ROM), 12 Pages, 2005/10
Safety demonstration tests using the HTTR are in progress to verify the inherent safety features, to improve the safety design and the technologies for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely becomes a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The SIRIUS code was developed to analyze reactor transient during the tests with reactor dynamics. This paper describes the validation of the SIRIUS code with the measured values of one and two gas circulators tripping test at 30% (9 MW). It was confirmed that the SIRIUS code was able to analyze the reactor transient within 10% during the tests. The result of this study and the way of resolving problems can be applied to development for not only the commercial HTGRs but also the Very High Temperature Reactor (VHTR) as one of the Generation IV reactors.
Naka Fusion Research Establishment
JAERI-Review 2005-046, 113 Pages, 2005/09
JAERI-Review-2005-046.pdf:24.35MB
This annual report provides an overview of research and development activities at Naka Fusion Research Establishment, including those performed in collaboration with other research establishments of JAERI, research institutes, and universities, during the period from 1 April, 2004 to 31 March, 2005. The activities in the Naka Fusion Research Establishment are highlighted by researches in JT-60 and JFT-2M, theoretical and analytical plasma researches, research and development of fusion reactor technologies towards ITER and fusion power demonstration plants, and activities in support of ITER design and construction.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
JAERI-Data/Code 2005-003, 31 Pages, 2005/06
JAERI-Data-Code-2005-003.pdf:4.83MB
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.
Ono, Tomio*; Subekti, M.*; Kudo, Kazuhiko*; Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Nabeshima, Kunihiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.115 - 126, 2005/06
Control-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs. This paper describes pre-test analysis method using artificial neural networks to predict the changes of reactor power and reactivity. The network model applied in this study is based on recurrent neural networks. The inputs of the network are the changes of the central control rods position and other significant core parameters, and the outputs are the changes of reactor power and reactivity. Furthermore, Time Synchronizing Signal(TSS) is added to input to improve the modeling of time series data. The actual tests data, which were previously carried out in the HTTR, were used for learning the model of the plant dynamics. After the learning, the network can predict the changes of reactor power and reactivity in the following tests.
Al superconductors for a fusion demo plantKoizumi, Norikiyo; Takeuchi, Takao*; Okuno, Kiyoshi
Nuclear Fusion, 45(6), p.431 - 438, 2005/06
Times Cited Count：32 Percentile：69.42(Physics, Fluids & Plasmas)no abstracts in English
Nakagawa, Shigeaki; Sakaba, Nariaki; Takamatsu, Kuniyoshi; Takada, Eiji*; Tochio, Daisuke; Owada, Hiroyuki*
JAERI-Tech 2005-015, 26 Pages, 2005/03
JAERI-Tech-2005-015.pdf:1.77MB
Safety demonstration tests using the HTTR are in progress since 2002 to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to not only the commercial HTGRs but also the research and development for the VHTR one of the Generation IV reactor candidates. This paper describes the reactivity insertion test (SR-3), the coolant flow reduction test by tripping of gas circulators (S1C-3/S2C-3), and the partial flow loss of coolant test (SF-2) planned in March 2005 with their detailed test method, procedure and results of pre-test analysis. From the analytical results, it was found that the negative reactivity feedback effect of the core brings the reactor power safely to a stable level without a reactor scram.
Tachibana, Yukio
Genshiryoku Nenkan 2005-Nen Ban, p.279 - 287, 2005/00
no abstracts in English
