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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:2 Percentile:25.63(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.

Journal Articles

Seismic analysis of the ITER blanket remote handling system

Noguchi, Yuto; Maruyama, Takahito; Takeda, Nobukazu; Kakudate, Satoshi

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

This paper reports the seismic analysis of the ITER Blanket RH system (BRHS) during blanket module handling operation. Since the BRHS, which is composed of the articulated rail and the vehicle manipulator, which travels on the rail deployed in the vacuum vessel in the toroidal direction, has various configurations and the rail system has flexibility, evaluation of dynamic response of the system is of essential importance. Via parameter sensitivity study on position and posture of the vehicle manipulator, the most unfavorable configuration for each component of the BRHS has been specified by the modal and spectrum analyses with the global BRHS FE model. Then using the quasi-static equivalent loads on the individual components obtained by the global BRHS seismic analysis, the structural verifications of the structural members of the BRHS have been carried out with detailed partial FE models. The system seismic resistance of the BRHS to a safe shutdown earthquake was confirmed.

Journal Articles

Availability analysis of the ITER blanket remote handling system

Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

Plasma and Fusion Research (Internet), 10(Sp.2), p.3405010_1 - 3405010_4, 2015/02

Journal Articles

Preliminary assessment for dust contamination of ITER in-vessel transporter

Saito, Makiko; Ueno, Kenichi; Maruyama, Takahito; Murakami, Shin; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka*; Tesini, A.*

Fusion Engineering and Design, 89(9-10), p.2352 - 2356, 2014/10

 Times Cited Count:7 Percentile:55.77(Nuclear Science & Technology)

After plasma operation of the ITER reactor, irradiated radioactive dust will accumulate in the vacuum vessel (VV). The In Vessel Transporter (IVT) will be installed in the VV and remove the blanket modules for maintenance. The IVT will be carried back to the Hot Cell Facilities (HCF) after exchanging the blanket, and the IVT itself also needs maintenance. It is considered that the maintenance workers will be exposed to the irradiated radioactive dust attached to the IVT surface. In this study, dust contamination of the IVT is evaluated to assess exposure during maintenance work in the HCF. The IVT contamination scenario is assumed in the ITER project. From plasma shut down until maintenance is performed on the IVT will take 345 days under the ITER project assumption. Under this scenario, the effective dose rate from irradiated radioactive dust was calculated as an infinite plate for each nuclide. As a result, W-181 and Ta-182 were the dominant nuclides for the effective dose rate. If all dust is W-181 or Ta-182, the effective dose rate is about 400 $$mu$$Sv/h and 100 $$mu$$Sv/h respectively. Nevertheless, using the dose limit determined by the ITER project and the estimated maximum maintenance time, the effective dose rate limit was calculated to be 4.18 $$mu$$Sv/h under these limited conditions. To satisfy the dose rate limit, decontamination processes were assumed and the dose rate after decontamination was evaluated.

Journal Articles

Robot vision system R&D for ITER blanket remote-handling system

Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, A.*

Fusion Engineering and Design, 89(9-10), p.2404 - 2408, 2014/10

 Times Cited Count:4 Percentile:36.81(Nuclear Science & Technology)

Journal Articles

Dust removal experiments for ITER blanket remote handling system

Ueno, Kenichi; Aburadani, Atsushi; Saito, Makiko; Maruyama, Takahito; Takeda, Nobukazu; Murakami, Shin; Kakudate, Satoshi

Plasma and Fusion Research (Internet), 9, p.1405012_1 - 1405012_4, 2014/02

Journal Articles

Rail deployment operation test for ITER blanket handling system with positioning misalignment

Takeda, Nobukazu; Aburadani, Atsushi; Tanigawa, Hisashi; Shigematsu, Soichiro; Kozaka, Hiroshi; Murakami, Shin; Kakudate, Satoshi; Nakahira, Masataka; Tesini, A.*

Fusion Engineering and Design, 88(9-10), p.2186 - 2189, 2013/10

 Times Cited Count:2 Percentile:21.27(Nuclear Science & Technology)

R&D for rail deployment equipment was performed for the ITER blanket remote handling system. The target torque for the automatic operation was investigated. The result shows that the 20% of the rated torque is adequate as the torque limitation for the automatic operation. A schedule for the procurement of the blanket remote handling system, which will be delivered to the ITER in 2020, was also shown.

Journal Articles

Performance evaluation on force control for ITER blanket installation

Aburadani, Atsushi; Takeda, Nobukazu; Shigematsu, Soichiro; Murakami, Shin; Tanigawa, Hisashi; Kakudate, Satoshi; Nakahira, Masataka*; Hamilton, D.*; Tesini, A.*

Fusion Engineering and Design, 88(9-10), p.1978 - 1981, 2013/10

 Times Cited Count:2 Percentile:21.27(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Irradiation test progress for the ITER maintenance robot

Noguchi, Yuto; Anzai, Katsunori; Kozaka, Hiroshi; Aburadani, Atsushi; Kazawa, Minoru; Takeda, Nobukazu; Kakudate, Satoshi

Dai-31-Kai Nihon Robotto Gakkai Gakujutsu Koenkai Yokoshu (DVD-ROM), 2 Pages, 2013/09

BB2013-1301.pdf:0.22MB

Journal Articles

Comparative study of laser and TIG weldings for application to ITER blanket hydraulic connection

Tanigawa, Hisashi; Aburadani, Atsushi; Shigematsu, Soichiro; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Jokinen, T.*; Merola, M.*

Fusion Engineering and Design, 87(7-8), p.999 - 1003, 2012/08

 Times Cited Count:13 Percentile:74.08(Nuclear Science & Technology)

This paper presents results of R&D activities where the laser and TIG welding tools were developed to apply the blanket hydraulic connection. The target pipe is 48.26 mm in outer diameter and 2.77 mm-thick. A single path welding without filler materials is required to reduce the weld heat input related to re-weldability. For the laser welding, the focal spot diameter was expanded to increase allowable misalignment. The TIG welding tool was equipped with AVC (Arc Voltage Control) to avoid a torch sticking and to enlarge allowable misalignment. For each tool, the welding conditions were optimized for all position welding to horizontally located pipes. Obtained parameters such as the weld heat input, allowable misalignment, lifetime of the tools and amount of sputter and fume, were comparatively assessed.

Journal Articles

Verification test results of a cutting technique for the ITER blanket cooling pipes

Shigematsu, Soichiro; Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Mori, Seiji*; Nakahira, Masataka*; Raffray, R.*; Merola, M.*

Fusion Engineering and Design, 87(7-8), p.1218 - 1223, 2012/08

 Times Cited Count:5 Percentile:42.85(Nuclear Science & Technology)

The current design of the ITER blanket system is a modular configuration and a total of 440 blanket modules are to be installed in the ITER vacuum vessel. Each blanket module consists of the first wall (FW) and the shield block (SB). The FW receives a high heat load from the plasma. The SB shields components from the neutrons generated by the nuclear fusion reaction. The FW will be damaged by the heat load and neutrons, so it requires scheduled replacement. For the FW replacement, cutting/welding tools for the cooling pipes must be able to conduct the following operations: access and cut/weld the pipe from the inside of the cooling pipe. The cutting tool for the pipe end is required to cut flat plate circularly from the surface side of the FW. This paper describes the current status of R&D of the cutting tools for maintenance of the cooling pipe of the FW.

Journal Articles

Maintenance concept for the SlimCS DEMO reactor

Tobita, Kenji; Uto, Hiroyasu; Kakudate, Satoshi; Takase, Haruhiko; Asakura, Nobuyuki; Someya, Yoji; Liu, C.

Fusion Engineering and Design, 86(9-11), p.2730 - 2734, 2011/10

 Times Cited Count:13 Percentile:74.14(Nuclear Science & Technology)

For high availability of DEMO operation, sector horizontal transport hot cell maintenance scheme was studied. Transport of sector with 730 tons is carried out using a wheeled platform. The driving force of pulling the sector into a cask is ball screws. The fulcrum of the ball screws is the cryostat wall so that a large pulling force is expected with no-counter balance. The cask containing the sector is delivered by air casters from the cryostat to the hot cell. For the maintenance scheme, new concepts such as transfer of the tilting forces of toroidal coils using ropes and shafts and supports for the tilting force using reinforced concrete floor or cryostat wall were proposed. Based on the maintenance concept, the period required for replacement of all sectors is estimated to be 35.5-67.5 days, satisfying the design target (shorter than 3 months).

Journal Articles

Operation and maintenance

Kakudate, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 87(Suppl.), p.194 - 200, 2011/02

no abstracts in English

Journal Articles

R&D on major components of control system for ITER blanket maintenance equipment

Takeda, Nobukazu; Kakudate, Satoshi; Matsumoto, Yasuhiro; Kozaka, Hiroshi; Aburadani, Atsushi; Negishi, Yusuke; Nakahira, Masataka*; Tesini, A.*

Fusion Engineering and Design, 85(7-9), p.1190 - 1195, 2010/12

 Times Cited Count:1 Percentile:11(Nuclear Science & Technology)

Several R&Ds for the ITER blanket remote handling system had been performed from the Engineering Design Activity phase until now and only several technical issues regarding the control system remained such as noise caused by slip ring, control of cable handling system, signal transmission through very long cable and radiation-hard amplifier. This study concentrates on these issues. As a conclusion, major issues for the control system have been solved and the ITER blanket remote handling system becomes further feasible.

JAEA Reports

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m$$^{2}$$. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Journal Articles

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

 Times Cited Count:125 Percentile:98.18(Physics, Fluids & Plasmas)

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m$$^{2}$$ or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

Journal Articles

Design progress of the ITER blanket remote handling equipment

Nakahira, Masataka; Matsumoto, Yasuhiro; Kakudate, Satoshi; Takeda, Nobukazu; Shibanuma, Kiyoshi; Tesini, A.*

Fusion Engineering and Design, 84(7-11), p.1394 - 1398, 2009/06

 Times Cited Count:19 Percentile:79.52(Nuclear Science & Technology)

Invessel components of ITER have to be maintained by remote handling (RH) equipment due to high radiation level in the vacuum vessel (VV) after D-D operation. Blanket module (BM) is maintained by a manipulator mounted on a vehicle traveled through an articulated rail deployed inside the VV. Towards the construction, the BLRH equipment design has been improved and developed in more detail. The overview of design results are introduced in this paper. The design of rail deployment system of the BLRH has been updated to enable the rail connection in the transfer cask in order to minimize occupation space. For this purpose, design works have been performed for concept, sequence and typical simulation of BL replacement in the VV and rail deployment of the RH equipment in the cask, including cask docking. The technical issues of the rail connection in the cask are (1) tight tolerance of a pin at a hinge, (2) limited space of the connection inside a cask and (3) tight positioning accuracy. This paper summarizes the idea to solve these issues and a result of the design work. The paper also introduces a new cable handling equipment, rail support equipment and BL receiver/transporter.

Journal Articles

Mock-up test on key components of ITER blanket remote handling system

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Ko; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, A.*

Fusion Engineering and Design, 84(7-11), p.1813 - 1817, 2009/06

 Times Cited Count:11 Percentile:63.13(Nuclear Science & Technology)

The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high $$gamma$$-ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The JAEA is continuing several R&Ds so that the system can be procured smoothly to ITER. The residual key issues after the EDA are rail connection, cable handling and in-situ replacement of first wall. The last issue is newly raised and currently under the discussion. This presentation concentrates on the former two issues.

Journal Articles

A Proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Onozuka, Masanori*

Fusion Engineering and Design, 83(10-12), p.1578 - 1582, 2008/12

 Times Cited Count:4 Percentile:31.93(Nuclear Science & Technology)

The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience.

Journal Articles

Development of a virtual reality simulator for the ITER blanket remote handling system

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, A.*

Fusion Engineering and Design, 83(10-12), p.1837 - 1840, 2008/12

 Times Cited Count:11 Percentile:56.05(Nuclear Science & Technology)

The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of $$gamma$$ ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of $$gamma$$ ray. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software "ENVISION". The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation.

131 (Records 1-20 displayed on this page)