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Journal Articles

Recent advances in modeling and simulation of the exposure and response of tungsten to fusion energy conditions

Marian, J.*; Becquart, C. S.*; Domain, C.*; Dudarev, S. L.*; Gilbert, M. R.*; Kurtz, R. J.*; Mason, D. R.*; Nordlund, K.*; Sand, A. E.*; Snead, L. L.*; et al.

Nuclear Fusion, 57(9), p.092008_1 - 092008_26, 2017/06

 Times Cited Count:56 Percentile:0.45(Physics, Fluids & Plasmas)

Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements. For plasma facing components such as the first wall and the divertor, tungsten (W) has been selected as the best candidate material due to its superior high-temperature and irradiation properties. In this paper we provide a review of recent efforts in computational modeling of W both as a plasma-facing material as well as a bulk structural material subjected to fast neutron irradiation. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining future challenges.

Journal Articles

Impact hammer test of ITER blanket remote handling system

Noguchi, Yuto; Maruyama, Takahito; Ueno, Kenichi; Komai, Masafumi; Takeda, Nobukazu; Kakudate, Satoshi

Fusion Engineering and Design, 109-111(Part B), p.1291 - 1295, 2016/11

 Times Cited Count:1 Percentile:82.26(Nuclear Science & Technology)

This paper reports the impact hammer test of the full-scale mock-up of ITER Blanket Remote Handling system (BRHS). Since the BRHS, which is composed of the articulated rail and the vehicle manipulator which travels on the rail deployed in the vacuum vessel, is subjected to the floor response spectrum with 14 G peak at 8 Hz, evaluation of dynamic response of the system is of essential importance. Recently impact hammer testing on the full-scale mock-up of the BRHS was carried out to verify the finite element method seismic analysis and to experimentally obtain the damping ratio of the system. The results showed that the mock-up has a vertical major natural mode with a natural frequency of 7.5 Hz and a damping ratio of 0.5%. While higher structural damping ratios is predicted in a high amplitude excitation such as major earthquake, it was confirmed that the experimental natural major frequencies are in agreement with the major frequencies obtained by elastic dynamic analysis.

JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (In-advance evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-002, 40 Pages, 2016/03

JAEA-Evaluation-2016-002.pdf:2.66MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for in-advance evaluation of "Research and Development of the technical system for extraction of fusion energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2015 to March 2022. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

JAEA Reports

Assessment report of research and development activities in FY2014; Activity "Fusion Research and Development" (Result evaluation)

Fusion Research and Development Directorate

JAEA-Evaluation 2016-001, 128 Pages, 2016/03

JAEA-Evaluation-2016-001.pdf:33.25MB

Japan Atomic Energy Agency (hereinafter referred to as "JAEA") asked the assessment committee, "Evaluation Committee of Research and Development Activities for Fusion" (hereinafter referred to as "Committee") for result evaluation of "Research and Development of the Technical System for Extraction of Fusion Energy," in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology " and "Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to the JAEA's request, the Committee assessed the research program of the Fusion Research and Development Directorate (hereinafter referred to as "FRDD") during the period from April 2010 to November 2014. The Committee evaluated the management and research activities of the FRDD based on the explanatory documents prepared by the FRDD, the oral presentations with questions-and-answers by the Director General and the Deputy Director Generals.

Journal Articles

Development of the negative ion beams relevant to ITER and JT-60SA at Japan Atomic Energy Agency

Hanada, Masaya; Kojima, Atsushi; Tobari, Hiroyuki; Nishikiori, Ryo; Hiratsuka, Junichi; Kashiwagi, Mieko; Umeda, Naotaka; Yoshida, Masafumi; Ichikawa, Masahiro; Watanabe, Kazuhiro; et al.

Review of Scientific Instruments, 87(2), p.02B322_1 - 02B322_4, 2016/02

 Times Cited Count:10 Percentile:37.37(Instruments & Instrumentation)

In International Thermo-nuclear Experimental Reactor (ITER) and JT-60 Super Advanced (JT-60 SA), the D$$^{-}$$ ion beams of 1 MeV, 40 A and 0.5 MeV, 22 A are required to produce 3600 s and 100 s for the neutral beam injection, respectively. In order to realize such as powerful D$$^{-}$$ ion beams for long duration time, Japan Atomic Energy Agency (JAEA) has energetically developed cesium (Cs)-seeded negative ion sources (CsNIS) and electro-static multi-aperture and multi-stage accelerators (MAMuG accelerator) which are chosen as the reference design of ITER and JT-60 SA. In the development of the CsNIS, a 100s production of the H$$^{-}$$ ion beam has been demonstrated with a beam current of 15 A by modifying the JT-60 negative ion source. At the higher current, the long pulse production of the negative ions has been tried by the mitigation of the arcing in the plasma inside the ion source. As for the long pulse acceleration of the negative ions in the MAMuG accelerator, the beam steering angle has been controlled to reduce the power loading of the acceleration grids A pulse duration time has been significantly extended from 0.4 s to 60 s at reasonable beam power for ITER requirement. The achieved pulse duration time is limited by the capacity of the power supplies in the test stand. In the range of $$<$$ 60 s, there are no degradations of beam optics and voltage holding capability in the accelerator. It leads to the further extension of the pulse duration time at higher power density. This paper reports the latest results of development on the negative ion source and accelerator at JAEA.

Journal Articles

Thermohydraulic responses of a water-cooled tokamak fusion DEMO to loss-of-coolant accidents

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*

Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12

 Times Cited Count:5 Percentile:64.1(Physics, Fluids & Plasmas)

Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.

Journal Articles

Fabrication and hydrogen generation reaction with water vapor of prototypic pebbles of binary beryllides as advanced neutron multiplier

Nakamichi, Masaru; Kim, Jae-Hwan

Fusion Engineering and Design, 98-99, p.1838 - 1842, 2015/10

 Times Cited Count:13 Percentile:13.04(Nuclear Science & Technology)

Advanced neutron multipliers with high stability at high temperatures are desired for the pebble bed blankets of DEMO reactors. Beryllium intermetallic compounds (beryllides) are the most promising material for this purpose. To fabricate the beryllide pebbles, a new granulation process has been established that combines a plasma sintering method for beryllide synthesis and a rotating electrode method using a plasma-sintered electrode for granulation. In granulation examinations, prototypic pebbles 1 mm in diameter of Be-V beryllide as well as Be-Ti beryllide were successfully fabricated. This study performed not only granulation of binary beryllides but also its characterization of the hydrogen generation reaction with water vapor compared with those of pure Be pebbles.

Journal Articles

Hydrophobic platinum honeycomb catalyst to be used for tritium oxidation reactors

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*; Noguchi, Hiroshi*; Edao, Yuki; Taniuchi, Junichi*

Fusion Science and Technology, 68(3), p.596 - 600, 2015/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

We have newly developed the hydrophobic platinum honeycomb catalysts applicable to tritium oxidation reactor since the honeycomb-shape catalyst can decrease the pressure drop. Two types of hydrophobic honeycomb catalyst have been test-manufactured. One is the hydrophobic platinum catalyst on a metal honeycomb. The other is the hydrophobic platinum catalyst on a ceramic honeycomb made of silicon carbide. The fine platinum particles around a few nanometers significantly improve the catalytic activity for the oxidation tritium at a tracer concentration. The hydrogen concentration in the gaseous feed slightly affects the overall reaction rate constant for hydrogen oxidation. Due to the competitive adsorption of hydrogen and water molecules on platinum surface, the overall reaction rate constant has the bottom value. The hydrogen concentration for the bottom value is 100 ppm under the dry feed gas. We have experimentally confirmed the activity of these honeycomb catalysts is as good as that of pellet-shape hydrophobic catalyst. The results support the hydrophobic honeycomb catalysts are applicable to tritium oxidation reactor.

Journal Articles

Catalyst technology of Tanaka Kikinzoku Kogyo

Kubo, Hitoshi*; Oshima, Yusuke*; Iwai, Yasunori

JETI, 63(10), p.33 - 36, 2015/09

Tanaka Kikinzoku Kogyo provides a broad range of precious metals products and technologies. Tanaka Kikinzoku Kogyo and Japan Atomic Energy Agency have jointly developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530 kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water. It is also anticipated that the hydrophobic platinum catalyst manufacturing technology used for this catalyst could be applied to a wide range of fields other than nuclear fusion research. It was verified that if applied to a hydro oxidation catalyst, hydrogen could be efficiently oxidized, even at room temperature. This catalyst can also contribute to improving safety at non-nuclear plants that use hydrogen in general by solving the aforementioned vulnerability issue.

Journal Articles

Development of hydrophobic platinum catalyst for the effective collection of tritium in fusion plants

Iwai, Yasunori; Kubo, Hitoshi*; Oshima, Yusuke*

Isotope News, (736), p.12 - 17, 2015/08

We have successfully developed a new hydrophobic platinum catalyst for collecting tritium at nuclear fusion reactors. Catalysts used to collect tritium are called hydrophobic precious metal catalysts. In Japan, hydrophobic precious metal catalysts manufactured from polymers have been used for heavy water refinement.However, this catalyst has issues related to embrittlement to radiation and thermal stability. These technological issues needed to be solved to allow for its application to nuclear fusion reactors requiring further enrichment from highly-concentrated tritiated water. We developed a new method of manufacturing catalysts involving hydrophobic processing with an inorganic substance base. As a result, previous technological issues were able to be solved with the development of a catalyst that exhibited no performance degradation in response to radiation application of 530kGy, a standard for radiation resistance, and maintenance of thermal stability at over 600$$^{circ}$$C, which is much higher than the 70$$^{circ}$$C temperature that is normally used. The catalyst created with this method was also confirmed to have achieved the world's highest exchange efficiency, equivalent to 1.3 times the previously most powerful efficiency. The application of this catalyst to the liquid phase catalytic exchange process is expected to overcome significant technological hurdles with regards to improving the reliability and efficiency of systems for collecting tritium from tritiated water.

Journal Articles

SOL-divertor plasma simulations introducing anisotropic temperature with virtual divertor model

Togo, Satoshi*; Takizuka, Tomonori*; Nakamura, Makoto; Hoshino, Kazuo; Ogawa, Yuichi*

Journal of Nuclear Materials, 463, p.502 - 505, 2015/08

 Times Cited Count:8 Percentile:27.18(Materials Science, Multidisciplinary)

A 1D SOL-divertor plasma simulation code introducing the anisotropic ion temperature with virtual divertor model has been developed. By introducing the anisotropic ion temperature directly, the second-derivative parallel ion viscosity term in the momentum transport equation can be excluded and the boundary condition at the divertor plate becomes unnecessary. In order to express the effects of the divertor plate and accompanying sheath implicitly, the virtual divertor model has been introduced which has an artificial sinks of particle, momentum and energy. The virtual divertor model makes the periodic boundary condition available. By using this model, SOL-divertor plasmas satisfying the Bohm condition has been successfully obtained. Also investigated are the dependence of the ion temperature anisotropy on the normalized mean free path of ion and the validity of the approximated parallel ion viscosity for the Braginskii expression and the limited one.

Journal Articles

Perspective of negative triangularity tokamak as fusion energy system

Kikuchi, Mitsuru; Medvedev, S.*; Takizuka, Tomonori*; Fasoli, A.*; Wu, Y.*; Diamond, P. H.*; Duan, X.*; Kishimoto, Yasuaki*; Hanada, Kazuaki*; 41 of others*

Europhysics Conference Abstracts (Internet), 39E, p.P4.179_1 - P4.179_4, 2015/06

Power and particle control in fusion reactor is quite a challenge and we have studied the negative triangularity tokamak (NTT) as an innovative concept to reduce the transient ELM heat load and the quasi steady-state heat load. A double-null NTT is stable to ideal MHD modes for a reactor relevant bN $$>$$ 3 while it is a magnetic hill configuration. In this paper, we report the configuration study of single-null NTT and its ideal MHD stability.

Journal Articles

Outreach activities in JAEA

Kasugai, Atsushi

Purazuma, Kaku Yugo Gakkai-Shi, 91(2), p.125 - 131, 2015/02

no abstracts in English

Journal Articles

Conference report; The 16th International Workshop on Radiative Properties of Hot Dense Matter

Sasaki, Akira; Nishimura, Hiroaki*; Onishi, Naofumi*

Purazuma, Kaku Yugo Gakkai-Shi, 91(2), p.166 - 167, 2015/02

no abstracts in English

Journal Articles

Fusion; The Energy of the Universe

McCracken, G.*; Stott, P.*; Iiyoshi, Atsuo*; Muraoka, Katsunori*; Nakai, Sadao*; Shimada, Michiya

Kaku Yugo; Uchu No Enerugi O Watashitachi No Te Ni, 326 Pages, 2015/00

Fusion; The Energy of the Universe, 2e is an essential reference providing basic principles of fusion energy from its history to the issues and realities progressing from the present day energy crisis. The book provides detailed developments and applications for researchers entering the field of fusion energy research. This second edition includes the latest results from the National Ignition Facility at the Lawrence Radiation Laboratory at Livermore, CA, and the progress on the International Thermonuclear Experimental Reactor (ITER) tokamak programme at Caderache, France.

Journal Articles

Actively convected liquid metal divertor

Shimada, Michiya; Hirooka, Yoshihiko*

Nuclear Fusion, 54(12), p.122002_1 - 122002_7, 2014/12

 Times Cited Count:22 Percentile:15.91(Physics, Fluids & Plasmas)

Tungsten is considered to be the most promising material for divertor in a fusion reactor. Tungsten divertor can withstand the heat loads of ITER, but the heat loads of DEMO divertor is a challenge. Pulsive heat loads as those associated with disruption could melt tungsten targets. The surface would not be flat after subsequent resolidification, which would significantly deteriorate its heat handling capability. Furthermore, DBTT of tungsten is rather high: $$sim$$ 400$$^{circ}$$C, which would become even higher after neutron irradiation, possibly resulting in cracks in tungsten. Our proposal is to use liquid metal for the divertor target material and actively circulate it with $$j$$ $$times$$ $$B$$ force. A simplified analysis of mhd equation in a cylindrical geometry suggests that the engineering requirement is modest. This analysis suggests that this new divertor concept merits further investigation.

Journal Articles

Status of development of Lithium Target Facility in IFMIF/EVEDA project

Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Hoashi, Eiji*; Fukada, Satoshi*; Suzuki, Akihiro*; Yagi, Juro*; Tsuji, Yoshiyuki*; et al.

Proceedings of Plasma Conference 2014 (PLASMA 2014) (CD-ROM), 2 Pages, 2014/11

In the IFMIF/EVEDA (International Fusion Materials Irradiation Facility/ Engineering Validation and Engineering Design Activity), the validation tests of the EVEDA lithium test loop with the world's highest flow rate of 3000 L/min was succeeded in generating a 100 mm-wide and 25 mm-thick free-surface lithium flow steadily under the IFMIF operation condition of a high-speed of 15 m/s at 250$$^{circ}$$C in a vacuum of 10 $$^{-3}$$ Pa. Some excellent results of the recent engineering validations including lithium purification, lithium safety, and remote handling technique were obtained, and the engineering design of lithium facility was also evaluated. These results will advance greatly the development of an accelerator-based neutron source to simulate the fusion reactor materials irradiation environment as an important key technology for the development of fusion reactor materials.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.

Journal Articles

Present status of the high current prototype accelerator for IFMIF/EVEDA

Okumura, Yoshikazu

Proceedings of 11th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.194 - 198, 2014/10

International Fusion Materials Irradiation Facility (IFMIF) is one of the major facilities in fusion program to irradiate and characterize the fusion materials under intense neutron field. Under the framework of Broader Approach (BA) agreement between Japan and Euratom, Engineering Validation and Engineering Design Activities (EVEDA) launched in 2007 in Japan to validate the key technologies to realize IFMIF. The most critical technology to realize IFMIF is two set of linear accelerators each producing 125mA/CW deuterium ion beams up to 40MeV. The prototype accelerator being developed in EVEDA consists of an injector, a RFQ accelerator, and a superconducting linac, whose target is to produce 125mA/CW deuterium ion beams up to 9MeV. The injector has been delivered to International Fusion Energy Research Center (IFERC) in Rokkasho, Japan, where all the accelerator components will be assembled and operated to validate the accelerator technologies required for IFMIF.

Journal Articles

Analysis of accident scenarios of a water-cooled tokamak DEMO

Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*

Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10

Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.

1157 (Records 1-20 displayed on this page)