Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Saito, Toru*; Morimoto, Masaaki*; Inagaki, Takashi*; Hong, Y.-S.*; Matsui, Kunihiro; Hemmi, Tsutomu; Kajitani, Hideki; et al.
Physics Procedia, 67, p.536 - 542, 2015/07
Japan Atomic Energy Agency (JAEA) has developed the tensile strength prediction method at liquid helium temperature (4K) using the quadratic curve as a function of the content of carbon and nitrogen in order to establish the rationalized quality control of the austenitic stainless steel used in the ITER superconducting coil operating at 4K. ITER is under construction aiming to verify technical demonstration of a nuclear fusion generation. Toroidal Field Coil (TFC), one of superconducting system in ITER, have been started procurement of materials in 2012. JAEA is producing materials for actual product which are the forged materials with shape of rectangle, round bar, asymmetry and etc. JAEA has responsibility to procure all ITER TFC Structures. In this process, JAEA obtained many tensile strength of both room temperature and 4K about these structural materials, for example, JJ1: High manganese stainless steel for structure (0.03C-12Cr-12Ni-10Mn-5Mo- 0.24N) and 316LN: High nitrogen containing stainless steel (0.2Nitrogen). Based on these data, accuracy of 4K strength prediction method for actual TFC Structure materials was evaluated and reported in this study.
Iguchi, Masahide; Sakurai, Takeru; Nakahira, Masataka; Koizumi, Norikiyo; Nakajima, Hideo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Application of partial penetration welding (PPW) to ITER Toroidal Field Coil structure has been proposed because of limited accessability for weld due to complex geometry and low stress and low importance components. In order to obtain fatigue crack growth (FCG) behavior of PPW joint in cryogenic environment, Japan Atomic Energy Agency performed FCG test at 4K by using Compact Tension (CT) specimens having as-weld notch of PPW. These CT specimens were made from mockups having one of actual joint shape of PPW, double J-groove. As the result of this test, it was observed that crack propagated in weld metal having inclination from as-weld notch. Moreover it was shown that FCG rate of as-weld CT specimens had high FCG rate region in early stage of crack propagation due to residual stress distribution. In addition, application method of this FCG rate to designing of PPW joint was proposed and verified in this study.
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
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 Sv/h and 100 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 Sv/h under these limited conditions. To satisfy the dose rate limit, decontamination processes were assumed and the dose rate after decontamination was evaluated.
Maruyama, Takahito; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Tesini, A.*
Fusion Engineering and Design, 89(9-10), p.2404 - 2408, 2014/10
Oshikawa, Takumi*; Funakoshi, Yoshihiko*; Imaoka, Hiroshi*; Yoshikawa, Kohei*; Maari, Yasutaka*; Iguchi, Masahide; Sakurai, Takeru; Nakahira, Masataka; Koizumi, Norikiyo; Nakajima, Hideo
Proceedings of 19th International Forgemasters Meeting (IFM 2014), p.254 - 259, 2014/09
ITER is a large-scale experiment that aims to demonstrate that it is possible to produce commercial energy from fusion. ITER Toroidal Field Coil Case (hereinafter referred to as "ITER TFCC") is one of the important components of ITER. The ITER TFCC materials are made of high nitrogen austenitic stainless steel and having various configurations. The ITER TFCC material which manufactured by JCFC has a complex configuration with heaver thickness than other materials. It is difficult to form near net shape to delivery configuration by ordinary open die forging method such as upset and stretching, because the ITER TFCC materials manufactured by JCFC have a complex configuration. Therefore ingot weight and lead time of machining increase when ITER TFCC materials are forged by ordinary open die forging method. Moreover, in order to get good attenuation at Ultrasonic examination, it is necessarily to make fine and uniform grain of the material. However, it is impossible to control grain size of austenitic stainless steel by heat treatment. The grain becomes fine and uniform by only forging process with suitable condition. Therefore, JCFC has studied suitable forging method to become near net shape to delivery configuration and also to get fine grain of center of the material. Based on these result, ITER TFCC materials were manufactured. This innovative forging process led to reduce the weight of ingot compared with general forging. And it had good Ultrasonic attenuation. It was confirmed that the results of material test and nondestructive examination satisfied the requirements of Japan domestic agency (hereinafter referred to as "JADA"). Moreover, the test coupons were taken from center of thick part of product and used for various tests. As the result of tests, it was confirmed that results of material test satisfied the requirements of JADA. It is clear that this innovative forging method is very suitable process for manufacturing of ITER TFCC materials.
Iguchi, Masahide; Morimoto, Masaaki; Chida, Yutaka*; Hemmi, Tsutomu; Nakajima, Hideo; Nakahira, Masataka; Koizumi, Norikiyo; Yamamoto, Akio*; Miyake, Takashi*; Sawa, Naoki*
IEEE Transactions on Applied Superconductivity, 24(3), p.3801004_1 - 3801004_4, 2014/06
no abstracts in English
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
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.
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
no abstracts in English
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
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.
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
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.
Nakahira, Masataka*; Niimi, Kenichiro; Irie, Hirosada*
Proceedings of 2009 ASME Pressure Vessels and Piping Division Conference (PVP 2009) (CD-ROM), 7 Pages, 2009/07
A standard of fabrication, installation, NDE and testing for superconducting magnets for fusion facility was developed. For construction of TF coil, a "Fabrication and Installation" standard FM-4000, accompanying a mandatory Appendix 41 "Welded Joint" and a "Nondestructive Examination" standard FM-5000 accompanying a mandatory Appendix 51 for "Ultrasonic Examination Method" and a "Pressure and Leak Testing" standard FM-6000 have been developed, based on other JSME standards for nuclear power plant (JSME S NB1) and also ASME Sec.III ND, NF or Sec.VIII-div.2 Since TF coil structure does not include radioactive materials but is operated under high stress produced by high magnetic field, it is not safety-relevant-barrier. The requirements to construction should be relaxed comparer with a fission reactor.
Nakahira, Masataka; Matsumoto, Yasuhiro; Kakudate, Satoshi; Takeda, Nobukazu; Shibanuma, Kiyoshi; Tesini, A.*
Fusion Engineering and Design, 84(7-11), p.1394 - 1398, 2009/06
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.
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
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 -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.
Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Onozuka, Masanori*
Fusion Engineering and Design, 83(10-12), p.1578 - 1582, 2008/12
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.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, A.*
Fusion Engineering and Design, 83(10-12), p.1837 - 1840, 2008/12
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 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 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.
Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.
Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12
This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.
Kakudate, Satoshi; Takeda, Nobukazu; Nakahira, Masataka; Matsumoto, Yasuhiro; Shibanuma, Kiyoshi; Tesini, A.*
Fusion Engineering and Design, 83(10-12), p.1850 - 1855, 2008/12
The design of in-vessel transporter (IVT) including vehicle manipulator has been updated according to the design changes such as blanket segmentation and structure, taking account of the interface between modules and vehicle manipulator. In particular, the updated design of the vehicle manipulator and rail has been carried out in order to avoid the interference between modules and vehicle manipulator. According to the updated design, the vehicle manipulator has been reduced by about 30%, compared with the reference design. In parallel with design activities, the R&D to clarify the specifications of the IVT design in detail is also performed, i.e., simulation system to provide the visual information during maintenance, dry lubricant to prevent the lubricant oil from spreading in the VV. The rail connection and cable handling in the transfer cask, which are critical issues for IVT system, are under preparation of the demonstration tests to finalize the design of the IVT system. Connection of the rail joint and cable handling test facilities are planned and under fabrication now. These test facility will be installed by the end of March 2008, and the performance tests will be carried out from April 2008.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi
Purazuma, Kaku Yugo Gakkai-Shi, 84(2), p.100 - 107, 2008/02
After the maintenance operation campaign at JET in 1998, which was the first fully remote handling operations on a fusion Tokamak, the remote maintenance of the in-vessel components is getting important more and more. For the fusion power plant, in particular, it is one of most important factors which affect on economics of the plant. Also for the ITER, currently under construction, the maintenance for the blanket and divertor in the vacuum vessel is planned to be operated by the remote equipment. In order to realize this, many designs and element tests were have been performed. This paper describes about the current situation of the research and development on the remote maintenance for fusion components, referring mainly the ITER remote maintenance system an example.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi
Plasma and Fusion Research (Internet), 2, p.052_1 - 052_4, 2007/12
In the present paper, a Diamond-Like Carbon (DLC) coating was assumed as a candidate of solid lubricants for the transmission gears of the ITER blanket maintenance equipment instead of liquid lubricant. The seizure tests using "pin-on-disk" method were performed with the disks of SCM440 and SNCM420 coated by the soft, layered and hard DLC. All cases satisfied the required allowable contact pressure, 2 GPa, and lifetime, cycles and thus, the feasibility of the DLC coating was validated. Among the three types, the soft DLC showed the best performance.
Shimizu, Katsusuke*; Onozuka, Masanori*; Usui, Yukinori*; Urata, Kazuhiro*; Tsujita, Yoshihiro*; Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Omori, Junji; Shibanuma, Kiyoshi
Fusion Engineering and Design, 82(15-24), p.2081 - 2088, 2007/10
To confirm the manufacturing and assembly process of the ITER vacuum vessel (VV), a series of related tests has been conducted. (1) Using a full-scale partial mock-up, fabrication methods are to be examined to determine feasibility. (2) To simulate a series of field-joint assembly operations, a test stand was built. (3) To provide an appropriate shield gas supply on the back side of the outer shell during field-joint welding, three types of back-seal structures have been tested. (4) The applicability of UT methods for volumetric inspection has been investigated. (5) Applicability of Liquid Penetrant Testing as a surface examination for the VV interior surface (i.e. ultra-vacuum side) has been investigated.