Kobayashi, Tatsuya*; Ito, Kimitaka*; Ido, Takeshi*; Kamiya, Kensaku*; Ito, Sanae*; Miura, Yukitoshi; Nagashima, Yoshihiko*; Fujisawa, Akihide*; Inagaki, Shigeru*; Ida, Katsumi*
Scientific Reports (Internet), 7(1), p.14971_1 - 14971_8, 2017/11
Spatiotemporal evolutions of radial electric field and turbulence are measured simultaneously in the H-mode transition, which is a prototypical example of turbulence structure formation in high-temperature plasmas. In the dynamical phase where transport barrier is established abruptly, the time-space-frequency-resolved turbulent particle flux is obtained. Here we report the validation of the mechanism of transport barrier formation quantitatively. It is found that the particle flux is suppressed predominantly by reducing density fluctuation amplitude and cross phase between density fluctuation and potential fluctuation. Both radial electric field shear and curvature are responsible for the amplitude suppression as was predicted by theory. Turbulence amplitude reduction immediately responds to the growth of the radial electric field non-uniformity and saturates, while cross phase continuously approaches zero.
Furukawa, Tomohiro; Hirakawa, Yasushi; Kondo, Hiroo; Kanemura, Takuji
Nuclear Materials and Energy (Internet), 9, p.286 - 291, 2016/12
In order to exchange the components which received irradiation damage during the operation at the International Fusion Materials Irradiation Facility, the adhered lithium, which is partially converted to lithium compounds such as lithium oxide and lithium hydroxide, should be removed from the components. In this study, the dissolution experiments of lithium compounds (lithium nitride, lithium hydroxide, and lithium oxide) were performed in a candidate solvent, allowing the clarification of time and temperature dependence. Based on the results, a cleaning procedure for adhered lithium on the inner surface of the components was proposed.
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
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.
Kim, Jae-Hwan; Nakamichi, Masaru
Fusion Engineering and Design, 109-111(Part.B), p.1764 - 1768, 2016/11
Beryllium intermetallic compounds (beryllides) are the most promising advanced neutron multipliers in demonstration fusion power reactors. Advanced neutron multipliers are being developed by Japan and the EU as part of their BA activities. Beryllides are too brittle to fabricate into pebble- or rod-like shapes using conventional methods such as arc melting and hot isostatic pressing. To overcome this issue, we developed a new combined plasma sintering and rotating electrode method for the fabrication of beryllide rods and pebbles. By using these methods, preliminary synthesis of the ternary beryllide pebbles with different chemical compositions, BeTiV (x=0.0-1.0) was successful. Scanning electron microscopic observation revealed that grain size on the surface decreased while area fraction of Be phase on cross-section decreased as V amount increased. These decreases may be contributed by the fact that the chemical composition of the pebble was closely varied to single-phase BeV with no peritectic reaction as V amount increased while Be, BeTi and BeTi phases were formed with large grain due to peritecic reaction in the BeTi. This feature influenced to variation of reactivity with 1% water vapor at high temperature. It was concluded that weight gain as well as H generation decreased due to decreases of grain size as well as Be phase on the surface as V amount increased in BeTiV.
Shirai, Hiroshi; Barabaschi, P.*; Kamada, Yutaka; JT-60SA Team
Fusion Engineering and Design, 109-111(Part.B), p.1701 - 1708, 2016/11
The JT-60SA Project has shown steady progress toward the first plasma in 2019. JT-60SA is a superconducting tokamak designed to operate in the break-even conditions for a long pulse duration with a maximum plasma current of 5.5 MA. Design and fabrication of JT-60SA components shared by EU and Japan started in 2007. Assembly in the torus hall started in January 2013, and welding work of the vacuum vessel sectors is currently on going on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so forth were or are being delivered to Naka site for installation, assembly and commissioning. This paper gives technical progress on fabrication, installation and assembly of tokamak components and ancillary systems, as well as progress of JT-60SA Research Plan being developed jointly by EU and Japanese fusion communities.
Konno, Chikara; Sato, Satoshi; Ota, Masayuki; Kwon, Saerom; Ochiai, Kentaro
Fusion Engineering and Design, 109-111(Part.B), p.1649 - 1652, 2016/11
Recently we have examined KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 in more detail and we found out the following new problems on the KERMA factors and DPA cross section data. (1) NJOY bugs and incorrect nuclear data generated KERMA factors and DPA cross section data of no increase with decreasing neutron energy in low neutron energy. (2) Huge helium production data caused drastically large KERMA factors and DPA cross section data in low neutron energy. (3) It seemed that NJOY could not adequately process capture cross section data in File 6, not File 12-15. (4) KERMA factors with the kinematics method are not correct for nuclear data libraries without detailed secondary particle data (energy-angular distribution data). These problems should be resolved based on our study.
Edao, Yuki; Sato, Katsumi; Iwai, Yasunori; Hayashi, Takumi
Journal of Nuclear Science and Technology, 53(11), p.1831 - 1838, 2016/11
Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking all the possible extraordinary situations in addition to that in a normal condition carefully into consideration. We focused on the change in efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor and tritium is fed into a catalytic reactor at the same time. Our test results indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in detritiation performance due to tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor enables to poison the catalyst temporarily and degrades the detritiation performance. Our investigation indicated a saturated water vapor condition without hydrocarbons would be the dominant scenario to determine the amount of catalyst for the design of catalytic reactor of the detritiation system.
Wakai, Eiichi; Watanabe, Kazuyoshi*; Ito, Yuzuru*; Suzuki, Akihiro*; Terai, Takayuki*; Yagi, Juro*; Kondo, Hiroo; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; et al.
Plasma and Fusion Research (Internet), 11, p.2405112_1 - 2405112_4, 2016/11
The lithium target facility of IFMIF (International Fusion Materials Irradiation Facility) consists of target assembly, lithium main loop, lithium purification loops, the diagnostic systems, and remote handling system. Major impurities in the lithium loop are proton, deuterium, tritium, 7-Be, activated corrosion products and the other species (C, N, O). It is very important to remove nitrogen content in lithium loop during operation, in order to avoid the corrosion/erosion of the nozzle of lithium target for the stable lithium flow on the target assembly. Nitrogen in the lithium can be removed by N hot trap using Fe-5at%Ti alloy at temperatures from 400 to 600C. In this study, the specification and the detailed design were evaluated, and the component of N hot trap system was fabricated.
Konno, Chikara; Kwon, Saerom; Ota, Masayuki; Sato, Satoshi
JAEA-Conf 2016-004, p.233 - 238, 2016/09
We compared the KERMA factors and DPA cross section data included in the official ACE and MATXS files of JENDL-4.0 with those of ENDF/B-VII.1 and JEFF-3.2. As a result, they were different from those of ENDF/B-VII.1 and JEFF-3.2 in a lot of nuclei, which was considered to be caused by the following new problems; (1) NJOY bugs, (2) huge helium production cross section data, (3) production data format in the nuclear data, (4) no detailed secondary particle data (energy-angular distribution data). The ACE and MATXS files of JENDL-4.0 with these problems should be revised based on this study.
Konno, Chikara; Kwon, Saerom; Ota, Masayuki; Sato, Satoshi
JAEA-Conf 2016-004, p.239 - 242, 2016/09
In order to specify reasons of the discrepancy between the calculated and measured results in analyses of benchmark experiments, some parts of some isotope data in nuclear data files are often modified and the modifies nuclear data files are processed with the NJOY code and the new ACE or MATXS files are used. However it is not easy to modify capture and elastic scattering data below 1 MeV with resonance data. Thus we devised a simple method to use capture and elastic scattering cross section data generated from resonance data with the NJOY code. This method was applied to detailed analyses of copper and molybdenum benchmark experiments at JAEA/FNS and it was demonstrated that this method was a very powerful tool.
Kobayashi, Tatsuya*; Ito, Kimitaka*; Ido, Takeshi*; Kamiya, Kensaku; Ito, Sanae*; Miura, Yukitoshi; Nagashima, Yoshihiko*; Fujisawa, Akihide*; Inagaki, Shigeru*; Ida, Katsumi*; et al.
Scientific Reports (Internet), 6, p.30720_1 - 30720_7, 2016/08
In this Letter, we report model verifications for a key physics for the L-H transition. Radial electric field excitation, using discharges having a two-step transition. Examining time derivative of Poissons equation, it is found that the sum of the loss cone loss current and neoclassical bulk viscosity current agrees the experimentally observed current in the order of magnitude at the first transition. Following second transition and an ambipolar condition in the L-mode cannot be explained by the models.
Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Sato, Satoshi; Kondo, Keitaro
Purazuma, Kaku Yugo Gakkai-Shi, 92(4), p.266 - 268, 2016/04
In the nuclear analysis of ITER, fusion DEMO reactor and IFMIF, etc., radiation transport calculations are conducted by using MCNP geometry data automatically created from 3-dimensional CAD data with CAD/MCNP conversion codes. In this review paper, we introduce the present status of the development on the CAD/MCNP conversion codes and examples of application. Also, we introduce how to visualize calculation results by MCNP.
Purazuma, Kaku Yugo Gakkai-Shi, 92(4), p.261 - 265, 2016/04
Useful information to not only beginners but also experts is introduced mainly for important points, which are basic but surprisingly unknown in, about calculation codes and nuclear data libraries in nuclear analyses for fusion reactors.
Oshima, Katsumi; Oda, Yasuhisa; Takahashi, Koji; Terakado, Masayuki; Ikeda, Ryosuke; Hayashi, Kazuo*; Moriyama, Shinichi; Kajiwara, Ken; Sakamoto, Keishi
JAEA-Technology 2015-061, 65 Pages, 2016/03
In JAEA, an ITER relevant control system for ITER gyrotron was developed according to Plant Control Design Handbook. This control system was developed based on ITER CODAC Core System and implemented state machine control of gyrotron operation system, sequential timing control of gyrotron oscillation startup, and data acquisition. The operation of ITER 170 GHz gyrotron was demonstrated with ITER relevant power supply configuration. This system is utilized for gyrotron operation test for ITER procurement. This report describes the architecture of gyrotron operation system, its basic and detailed design, and recent operation results.
Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Omori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; et al.
JAEA-Technology 2015-053, 36 Pages, 2016/03
The superconducting Satellite Tokamak machine "JT-60SA" under construction in Naka Fusion Institute is an international collaborative project between Japan (JA) and Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site. This means that Japan Atomic Energy Agency (JAEA) had a quite difficult issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the tough negotiation, and then the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. With the help of these approaches by JAEA, the EU works for QPC were successfully completed with no accident, and a great achievement was made for both EU and JA.
Kim, Jae-Hwan; Nakamichi, Masaru
JAEA-Review 2015-044, 260 Pages, 2016/03
The 12th International Workshop on Beryllium Technology (BeWS-12) was held on 10-11 September at the International Convention Center Jeju (ICCJEJU) in Jeju Island, Korea within a framework created by the International Energy Agency (IEA) Energy Technology Initiative on Fusion Materials. This workshop has been holding every 2 years from 1993. The objective of this workshop is to disseminate results of research and technology development in areas relevant to beryllium utilization in fusion nuclear power systems. In this workshop, a lot of researchers and technicians engaged in R&D on beryllium related materials and fusion engineering attended and discussed. This JAEA-Conf has been compiled the manuscripts and the presentation files in the BeWS-12.
Fusion Research and Development Directorate
JAEA-Evaluation 2016-002, 40 Pages, 2016/03
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.
Fusion Research and Development Directorate
JAEA-Evaluation 2016-001, 128 Pages, 2016/03
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.
Novello, L.*; Cara, P.*; Coletti, A.*; Gaio, E.*; Maistrello, A.*; Matsukawa, Makoto; Philipps, G.*; Tomarchio, V.*; Yamauchi, Kunihito
IEEE Transactions on Applied Superconductivity, 26(2), p.4700507_1 - 4700507_7, 2016/03
The voltage transient appearing across and inside the toroidal field (TF) coils of JT-60SA in case of fast voltage variation, such as a safety discharge operated by the quench protection circuit (QPC), can be significantly high. In fact, the voltage distribution between coils and inside the winding can be not uniform during fast transient, being influenced by the presence of parasitic capacitances. A simplified electrical model of the TF coils has been developed to investigate this aspect. The obtained model has been used in conjunction with an electrical model of the TF circuit elements, including a simplified model of the QPC. The worst case in terms of transient voltage applied to the winding has been identified, corresponding to a fault to ground occurring just after QPC operation. It has been verified that the resulting voltage is largely inside the coil insulation capability defined by performed insulation voltage tests.