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Journal Articles

Energy of the $$^{229}$$Th nuclear clock isomer determined by absolute $$gamma$$-ray energy difference

Yamaguchi, Atsushi*; Muramatsu, Haruka*; Hayashi, Tasuku*; Yuasa, Naoki*; Nakamura, Keisuke; Takimoto, Misaki; Haba, Hiromitsu*; Konashi, Kenji*; Watanabe, Makoto*; Kikunaga, Hidetoshi*; et al.

Physical Review Letters, 123(22), p.222501_1 - 222501_6, 2019/11

Journal Articles

The Present conditions of the acceleration voltage after the acceleration tube update

Matsuda, Makoto; Osa, Akihiko; Ishizaki, Nobuhiro; Tayama, Hidekazu; Nakanoya, Takamitsu; Kabumoto, Hiroshi; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro

JAEA-Conf 2018-003, p.126 - 131, 2019/02

no abstracts in English

Journal Articles

Gamma-ray spectrum from thermal neutron capture on gadolinium-157

Hagiwara, Kaito*; Yano, Takatomi*; Das, P. K.*; Lorenz, S.*; Ou, Iwa*; Sakuda, Makoto*; Kimura, Atsushi; Nakamura, Shoji; Iwamoto, Nobuyuki; Harada, Hideo; et al.

Progress of Theoretical and Experimental Physics (Internet), 2019(2), p.023D01_1 - 023D01_26, 2019/02

 Times Cited Count:1 Percentile:100(Physics, Multidisciplinary)

Journal Articles

Failure behavior analyses of piping system under dynamic seismic loading

Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*

International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11

 Times Cited Count:0 Percentile:100(Engineering, Multidisciplinary)

It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.

Journal Articles

Present status of JAEA-Tokai tandem accelerator

Matsuda, Makoto; Kabumoto, Hiroshi; Tayama, Hidekazu; Nakanoya, Takamitsu; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro; Matsui, Yutaka; Ishizaki, Nobuhiro; et al.

Proceedings of 15th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1271 - 1275, 2018/08

The JAEA-Tokai tandem accelerator was operated over a total of 64 days, and delivered 13 different ions to the experiments in the research fields of nuclear physics, nuclear chemistry, atomic physics, solid state physics and radiation effects in material in FY2017. After the vacuum accident occurred in December 2016 the accelerating voltage dropped to 12 MV. In order to remove dust and broken carbon foil in the accelerating tube, all 80 accelerator tubes were removed and rewashed. It took 4 months for cleaning and 2 months for reassembly. Therefore about 10 months were a maintenance period of an accelerator from February 2017. Along with the reconstruction of the accelerating tube, re-alignment of the accelerating tube was carried out. The operation resumed in December 2017 and it was possible to recover the maximum voltage to 17.4 MV without beam and 16.6 MV with beam with periodic conditioning work.

Journal Articles

Present status of JAEA-Tokai tandem accelerator

Kabumoto, Hiroshi; Osa, Akihiko; Ishizaki, Nobuhiro; Tayama, Hidekazu; Matsuda, Makoto; Nakanoya, Takamitsu; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro

Proceedings of 14th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1404 - 1408, 2017/12

The JAEA-Tokai tandem accelerator was operated over a total of 110 days, and delivered 22 different ions to the experiments in the research fields of nuclear physics, nuclear chemistry, atomic physics, solid state physics and radiation effects in material in FY2016. The damaged acceleration tubes by discharge were replaced with the spare tube at the regular maintenance period in March 2016, and the maximum accelerating voltage recovered to the 17 MV. However, an accident of vacuum breaking of all acceleration tubes was occurred in December, and accelerating voltage fell down to under 12 MV. Now, we are doing the maintenance to recover the performance of acceleration voltage. This paper describes the operational status of the accelerators and the major technical developments of our facility.

Journal Articles

Present status of JAEA-Tokai tandem accelerator and booster

Matsuda, Makoto; Osa, Akihiko; Ishizaki, Nobuhiro; Tayama, Hidekazu; Nakanoya, Takamitsu; Kabumoto, Hiroshi; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; Asozu, Takuhiro

Proceedings of 13th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.1413 - 1417, 2016/11

The tandem accelerator was operated over a total of 140 days and delivered 22 different ions to the experiments in the fields of nuclear physics, nuclear chemistry, atomic physics, solid state physics and radiation effects in material. Maximum accelerating voltage is keeping up 18 MV and there was used for ten days on this voltage. However, electric discharge was occurred frequently in December and accelerating voltage fell to 12 MV. The damaged acceleration tubes were replaced with the spare tube at the regular maintenance period in March. The superconducting booster was not operated. This paper describes the operational status of the accelerators and the major technical developments.

JAEA Reports

Summary of instructor training program in FY2014 aiming at Asian countries introducing nuclear technologies for peaceful use (Contract program)

Hidaka, Akihide; Nakano, Yoshihiro; Watanabe, Yoko; Arai, Nobuyoshi; Sawada, Makoto; Kanaizuka, Seiichi*; Katogi, Aki; Shimada, Mayuka*; Ishikawa, Tomomi*; Ebine, Masako*; et al.

JAEA-Review 2016-011, 208 Pages, 2016/07

JAEA-Review-2016-011-01.pdf:33.85MB
JAEA-Review-2016-011-02.pdf:27.68MB

JAEA has been conducting the Instructor Training Program (ITP) since 1996 under the auspices of MEXT to contribute to human resource development in currently 11 Asian countries in the field of radiation utilization for seeking peaceful use of nuclear energy. ITP consists of Instructor Training Course (ITC), Follow-up Training Course (FTC) and Nuclear Technology Seminars. In the ITP, trainings or seminars relating to technology for nuclear utilization are held in Japan by inviting nuclear related people from Asian countries to Japan and after that, the past trainees are supported during FTC by dispatching Japanese specialists to Asian countries. News Letter is also prepared to provide the broad range of information obtained through the trainings for local people near NPPs in Japan. The present report describes the activities of FY2014 ITP and future challenges for improving ITP more effectively.

Journal Articles

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

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.

JAEA Reports

Proceedings of the 21st Meeting of the International Collaboration on Advanced Neutron Sources (ICANS-XXI); Sep. 29 - Oct.3, 2014, Ibaraki Prefectural Center, Mito, Japan

Oku, Takayuki; Nakamura, Mitsutaka; Sakai, Kenji; Teshigawara, Makoto; Hideki, Tatsumoto*; Yonemura, Masao*; Suzuki, Junichi*; Arai, Masatoshi*

JAEA-Conf 2015-002, 660 Pages, 2016/02

JAEA-Conf-2015-002.pdf:168.34MB

The twenty first meeting of the International Collaboration on Advanced Neutron Source (ICANS-XXI) was held at Ibaraki Prefectural Culture Center in Mito from 29 September to 3 October 2014. It was hosted by Japan Atomic Energy Agency (JAEA), High Energy Accelerator Research Organization (KEK) and Comprehensive Research Organization for Science and Society (CROSS). In the meeting, new science and technology in the new era with the high power neuron sources were discussed in mostly "workshop style" sessions. In each session, various kinds of issues related to not only the hardware, but also the software and even radiation safety were discussed with the keyword of "INTERFACE". More than 200 Papers were presented in the meeting and 72 contributed papers are compiled in the proceedings.

Journal Articles

Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

Uto, Hiroyasu; Takase, Haruhiko; Sakamoto, Yoshiteru; Tobita, Kenji; Mori, Kazuo; Kudo, Tatsuya; Someya, Yoji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; et al.

Fusion Engineering and Design, 103, p.93 - 97, 2016/02

 Times Cited Count:7 Percentile:17.68(Nuclear Science & Technology)

Conceptual design of in-vessel component including conducting shell has been investigated in Broader Approach (BA) DEMO design activities, in order to propose feasible DEMO reactor from plasma vertical stability and engineering viewpoint. The conducting shell for the plasma vertical stability will be incorporated behind blanket module, while the location must be close to the plasma surface as possible for the plasma stabilization. We evaluated dependence of the plasma vertical stability on the conducing shell parameters by using a 3-dimensional eddy current analysis code (EDDYCAL). The calculation results showed that the conducting shell requires more than 0.01 m thickness of Cu-alloy on DEMO. On the other hand, the electromagnetic force at the plasma disruption is a few times larger than no conducting shell case because of larger eddy current on conducting shell. The engineering design issues of in-vessel components for plasma vertical stability are presented.

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:4 Percentile:66.55(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.

JAEA Reports

Basic Radiation Knowledge for School Education Course; Nuclear Technology Seminar 2014 (Contract program)

Watanabe, Yoko; Arai, Nobuyoshi; Sawada, Makoto; Kanaizuka, Seiichi; Shimada, Mayuka*; Ishikawa, Tomomi*; Nakamura, Kazuyuki

JAEA-Review 2015-026, 38 Pages, 2015/11

JAEA-Review-2015-026.pdf:10.55MB

JAEA has conducted Nuclear Technology Seminar for Asian countries which plan to introduce NPP, in order to increase the number of engineers and specialists. The Nuclear Technology Seminar on the Basic Radiation Knowledge for School Education Course was launched in 2012 due to increased recognition of the dissemination of the basic knowledge of radiation in public and education sectors as an important issue in the aftermath of the Fukushima Dai-ichi NPP Accident. In response to the requests of past participants, a new exercise "Joint experiment with high school students" was introduced from 2014 to provide an international learning experience for the course participants and the local Japanese students. A new learning material was also developed to help participants to study the basics of radiation in English. All the course activities including the details of preparatory process and course evaluation were described in this report.

Journal Articles

R&D activities of tritium technologies on Broader Approach in Phase 2-2

Isobe, Kanetsugu; Kawamura, Yoshinori; Iwai, Yasunori; Oyaizu, Makoto; Nakamura, Hirofumi; Suzuki, Takumi; Yamada, Masayuki; Edao, Yuki; Kurata, Rie; Hayashi, Takumi; et al.

Fusion Engineering and Design, 98-99, p.1792 - 1795, 2015/10

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

Activities on Broader Approach (BA) were started in 2007 on the basis of the Agreement between the Government of Japan and the EURATOM. The period of BA activities consist of Phase1 and Phase2 dividing into Phase 2-1 (2010-2011), Phase 2-2 (2012-2013) and Phase 2-3 (2014-2016). Tritium technology was chosen as one of important R&D issues to develop DEMO plant. R&D activities of tritium technology on BA consist of four tasks. Task-1 is to prepare and maintain the tritium handling facility in Rokkasho BA site in Japan. Task 2, 3 and 4 are main R&D activities for tritium and these are focused on: Task-2) Development of tritium accountancy technology, Task-3) Development of basic tritium safety research, Task-4) Tritium durability test. R&D activities of tritium technology in Phase 2-2 were underway successfully and closed in 2013.

Journal Articles

Design study of blanket structure based on a water-cooled solid breeder for DEMO

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Tokunaga, Shinsuke; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru

Fusion Engineering and Design, 98-99, p.1872 - 1875, 2015/10

 Times Cited Count:20 Percentile:3.62(Nuclear Science & Technology)

Blanket concept with simplified interior for mass production has been developed with a mixed bed of Li$$_{2}$$TiO$$_{3}$$ and Be$$_{12}$$Ti pebbles, a coolant condition of 15.5 MPa and 290-325$$^{circ}$$C and cooling tubes only without any partitions. A neutronics analysis ensured the blanket concept meets a self-sufficient supply of tritium. However, this concept is vulnerable to the inner pressure. A plant availability for DEMO may drop to a lower value, because a potential of resume operations after an accident such as a coolant leakage in blanket is not considered. The blanket design will be revisited for the availability. Considering the continuity with the ITER-TBM option of Japan and the engineering feasibility of fabrication, our design study focuses on a water-cooled solid breeding blanket using the mixed pebbles bed. A breakage of the blanket casing should be avoided not to contaminate the plasma chamber with water and breeding materials. A water-cooled solid blanket with inner pressure tightness is estimated by the ANSYS code. As a results, the pressure tightness of 8 MPa (water vapor pressure at 300$$^{circ}$$C) can be compatible with the self-sufficient production of tritium when the blanket is as thick as about 0.9 m and the ribs are arranged in the radial direction. Therefore, the blanket concept with pressure tightness of 8 MPa is adopted with depressurization system as which a tritium recovery system such as helium purge-gas line is posteriorly arranged in blanket to serve. On the other hand, a handling of decay heat is a serious problem at an accident such as LOCA. Coolant flow is divided into the blanket to secure heat removal for the safety. Finally, the blanket segmentation with the shape and dimension of blanket and routing of coolant flow has also been proposed. Moreover, overall TBR is estimated with torus configuration based in the segmentation using three-dimensional MCNP calculation.

Journal Articles

Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

Fusion Engineering and Design, 98-99, p.1648 - 1651, 2015/10

 Times Cited Count:7 Percentile:26.62(Nuclear Science & Technology)

Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field coil, the arrangement of poloidal field coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. In this study, we categorize various schemes in term of (1) the maintenance port position for transporting blanket segments, (2) blanket segmentation, and (3) divertor segmentation. In reviewing these assessment factors, the separated sector transport using the vertical maintenance ports with small divertor cassette maintenance scheme was found to be a more probable maintenance approach. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.

Journal Articles

Management strategy for radioactive waste in the fusion DEMO reactor

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke

Fusion Science and Technology, 68(2), p.423 - 427, 2015/09

 Times Cited Count:8 Percentile:22.06(Nuclear Science & Technology)

The radioactive waste is generated in every replacement of an in-vessel component. Maintenance scheme is to replace the blanket segment and divertor cassette independently, as the lifetime of them is different. The blanket segment consists of some blanket modules mounted to back-plate. Total weight is estimated to amount to about 6,648 ton (1,575 ton of blanket module, 3,777 ton of back-plate, 372 ton of conducting shell and 924 ton of divertor cassette). In base case, main parameters of DEMO reactor are 8.2 m of major radius and 1.35 GW of fusion output. The lifetimes of blanket segment and divertor cassette are assumed to be 2.2 years and 0.6 year, respectively, 52,487 ton wastes is generated in plant life of 20 years. Therefore, there is a concern that a contamination controlled area for the radioactive waste may increase because much the waste is generated in every replacement. In this paper, management scenario is proposed to reduce the radioactive waste. The back-plates and cassette bodies (628 ton) of divertor was reused. As a result, the displacement per atom (DPA) of the back-plates of SUS316L was 0.2 DPA/year and that of the cassette bodies of F82H was 0.6 DPA/year. Therefore, reusing the back-plates and cassette bodies would be possible, if re-welding points are arranged under neutron shielding. It was found that radioactive waste could be reduced to 20%, when tritium breeding materials are recycled. Finally, a design of DEMO building such as a hot cell and temporary storage etc. is proposed.

Journal Articles

Present status of JAEA-Tokai tandem accelerator and booster

Matsuda, Makoto; Osa, Akihiko; Abe, Shinichi; Ishizaki, Nobuhiro; Tayama, Hidekazu; Nakanoya, Takamitsu; Kabumoto, Hiroshi; Nakamura, Masahiko; Kutsukake, Kenichi; Otokawa, Yoshinori; et al.

Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.357 - 360, 2015/09

no abstracts in English

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:7 Percentile:26.62(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

Outlines of JAEA'S instructor training program and future prospects

Hidaka, Akihide; Nakamura, Kazuyuki; Watanabe, Yoko; Yabuuchi, Yukiko; Arai, Nobuyoshi; Sawada, Makoto; Yamashita, Kiyonobu; Sawai, Tomotsugu; Murakami, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

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