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JAEA Reports

Analytical study of cover plate welding deformation of the radial plate of the ITER toroidal field coil

Omori, Junji; Koizumi, Norikiyo; Shimizu, Tatsuya; Okuno, Kiyoshi; Hasegawa, Mitsuru*

JAEA-Technology 2009-046, 60 Pages, 2009/09

JAEA-Technology-2009-046.pdf:9.67MB

In the winding pack (WP) of the ITER TF coil, cover plates (CPs) are welded to radial plate (RP) after placing the conductors into the RP groove to fix it. The dimensions of the RP are 15 m high and 9 m wide, while its required tolerances are very severe such as flatness of 2 mm and in-plane deformation of about 2.5 mm. It is therefore necessary to reduce the deformation of the RP by CP welding. In order to estimate the weld deformation, a 1 m RP mock-up was fabricated and weld deformations were measured. From the test results, inherent strains have been obtained and the weld deformations of the full scale RP have been estimated. The RP deformations could be within the tolerances by the CP welding thickness of 2.5 mm in inboard region and 1.0 mm in outboard region. In addition, an alternative design, which improve the fabricability of the WP, was proposed. The analyses for the alternative design is performed and the results show the deformations could be reduced more.

Journal Articles

Engineering and maintenance studies of the ITER diagnostic upper port plug

Sato, Kazuyoshi; Omori, Junji; Kondoh, Takashi; Hatae, Takaki; Kajita, Shin*; Ishikawa, Masao; Neyatani, Yuzuru; Ebisawa, Katsuyuki*; Kusama, Yoshinori

Fusion Engineering and Design, 84(7-11), p.1713 - 1715, 2009/06

 Times Cited Count:1 Percentile:10.95(Nuclear Science & Technology)

Engineering analyses have been performed for the representative diagnostic upper port plug of ITER. Maintenance and integration design have been also carried out for the diagnostic components to be installed in the upper port plug. From the electromagnetic and structural analyses, it has come up an important problem to suppress the displacement of the upper port plug rather than to reduce the produced stress. Reducing the EM force will help to decrease the severity of potential displacement. Maximum displacement of the port plug decreases with increasing in the number of slits in a manner that the displacement would seem to be less than the design tolerance. A proposed low body roller and inner frame may enhance maintenance and integration. These studies and designs have established the design basis for the diagnostic upper port plug.

Journal Articles

Development of ITER diagnostic upper port plug

Sato, Kazuyoshi; Omori, Junji; Ebisawa, Katsuyuki*; Kusama, Yoshinori; Neyatani, Yuzuru

Plasma and Fusion Research (Internet), 2, p.S1088_1 - S1088_4, 2007/11

A part of diagnostic device in vacuum vessel is planned to install in the port plug to make sure the line of sight of diagnostics. Only basic concept is shown for the port plug since design of diagnostic devices has not been substantiated yet. The integration design of the port plug has been performed and the structure concept for electro-magnetic and neutron load has been investigated as for the No.11 upper port plug to confirm reliability of the proposed design. Three diagnostics will be installed in the No.11 upper port plug, the edge Thomson scattering system, the visible-IR TV divertor viewing system and the neutron activation system. To integrate theses diagnostic systems in the port plug, it was designed the arrangement of the labyrinth of optical path, the driving mechanism and cooling systems for shutters and mirrors, the maintenance space. The part just behind the blanket shield module (BSM) was changed to secure a space for maintenance and for associated diagnostic first mirror and shutter, whereas this place is assigned for neutron shielding in the present design. The BSM support, which is main component to apply the electro-magnetic load, was arranged with optical path inside BSM.

Journal Articles

Demonstration tests for manufacturing the ITER vacuum vessel

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

 Times Cited Count:5 Percentile:39.35(Nuclear Science & Technology)

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.

Journal Articles

Progress on design and development of ITER equatorial launcher; Analytical investigation and R&D of the launcher components for the design improvement

Takahashi, Koji; Kobayashi, Noriyuki; Omori, Junji; Suzuki, Satoshi; Kasugai, Atsushi; Sakamoto, Keishi

Fusion Science and Technology, 52(2), p.266 - 280, 2007/08

 Times Cited Count:24 Percentile:84.74(Nuclear Science & Technology)

Progress on design of an ITER equatorial EC Launcher with analytical and R&D studies of the components is described. The modified design of the front shield module is proposed with electromagnetic and structure analysis. The analytical investigation of the modified steering mirror design shows that maximum temperature and stress intensity are 289$$^{circ}$$C and 336 MPa on the mirror surface (copper alloy) and the inner surface of cooling tube (SS316) in the mirror, respectively. These results are less than allowable level. High heat irradiation experiment of the mirror mock-up and fatigue test of the spiral tube were carried out and their results convinced that the concept of the steering mirror structure was feasible. The results on neutron irradiation test of the composing material and the alternatives such as polyimide and liquid crystal polymer for an ultrasonic motor indicate that the motor with those materials is available for the ITER launcher.

Journal Articles

SlimCS; Compact low aspect ratio DEMO reactor with reduced-size central solenoid

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Nuclear Fusion, 47(8), p.892 - 899, 2007/08

 Times Cited Count:53 Percentile:87.5(Physics, Fluids & Plasmas)

The concept for a compact DEMO reactor named "SlimCS" is presented. Distinctive features of the concept is low aspect ratio ($$A$$ = 2.6) and use of a reduced-size center solenoid (CS) which has a function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field (TF) coil system which contributes to reducing the weight and construction cost of the reactor. SlimCS is as compact as advanced commercial reactor designs such as ARIES-RS and produces 1 GWe in spite of moderate requirements for plasma parameters. Merits of low-$$A$$, i.e. vertical stability for high elongation and high beta limit are responsible for such reasonable physics requirements.

JAEA Reports

Studies on representative disruption scenarios, associated electromagnetic and heat loads and operation window in ITER

Fujieda, Hirobumi; Sugihara, Masayoshi*; Shimada, Michiya; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji; Neyatani, Yuzuru

JAEA-Research 2007-052, 115 Pages, 2007/07

JAEA-Research-2007-052.pdf:3.58MB

Impacts of plasma disruptions on ITER have been investigated to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load. Heat load on the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code. For vertical displacement event, beryllium ($$Be$$) wall will not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper $$Be$$ wall and tungsten baffle due to the TQ after the vertical movement. However, its impact could be mitigated by implementing a reliable detection system of the vertical movement and a mitigation system.

Journal Articles

Disruption scenarios, their mitigation and operation window in ITER

Shimada, Michiya; Sugihara, Masayoshi; Fujieda, Hirobumi*; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

Several representative disruption scenarios are specified and disruption simulations are performed with the DINA code and EM load analyses with the 3D FEM code for these scenarios based on newly derived physics guidelines. Although some margin is confirmed in the EM loads due to induced eddy and halo currents on the in-vessel components for all of the representative scenarios, but the margin is not large. The heat load on various parts of the first wall due to vertical movements and thermal quenches is calculated. The beryllium wall will not melt during vertical movement. Melting is anticipated at the thermal quench during a VDE, though its impact could be reduced substantially by implementing a reliable detection and mitigation system, e.g., massive gas injection. With unmitigated disruptions, the loss of beryllium layer is expected to be within 30 $$mu$$m/event out of 10 mm thick beryllium first wall.

Journal Articles

Concept of compact low aspect ratio Demo reactor, SlimCS

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2006/10

no abstracts in English

JAEA Reports

Applicability assessment of plug weld to ITER vacuum vessel by crack propagation analysis

Omori, Junji; Nakahira, Masataka; Takeda, Nobukazu; Shibanuma, Kiyoshi; Sago, Hiromi*; Onozuka, Masanori*

JAEA-Technology 2006-017, 134 Pages, 2006/03

JAEA-Technology-2006-017.pdf:15.96MB

In order to improve the fabricability of the vacuum vessel (VV) of International Thermonuclear Experimental Reactor (ITER), applicability of plug weld between VV outer shell and stiffening ribs/blanket support housings has been assessed using crack propagation analysis for the plug weld. The ITER VV is a double-wall structure of inner and outer shells with ribs and housings between the shells. For the fabrication of VV, ribs and housings are welded to outer shell after welding to inner shell. A lot of weld grooves should be adjusted for the outer shell weld. The plug weld can allow larger tolerance of weld groove gaps than ordinary butt weld. However, un-welded lengths parallel to outer shell surface remain in the plug weld region. It is necessary to evaluate the allowable un-welded length to apply the plug weld to ITER vacuum vessel fabrication. For the assessment the allowable un-welded lengths have been calculated by crack propagation analyses for the load conditions, conservatively assuming the un-welded region is a crack. The analyses have been carried out for typical inboard straight region and inboard upper curved region with maximum housing stress. The allowable cracks of ribs are estimated to be 8.8mm and 38mm for the rib and the housing, respectively, considering inspection error of 4.4mm. Plug welding for welding between outer shell and ribs/housings could be applicable.

Journal Articles

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

 Times Cited Count:113 Percentile:99.15(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Consideration on blanket structure for fusion DEMO plant at JAERI

Nishio, Satoshi; Omori, Junji*; Kuroda, Toshimasa*; Tobita, Kenji; Enoeda, Mikio; Tsuru, Daigo; Hirose, Takanori; Sato, Satoshi; Kawamura, Yoshinori; Nakamura, Hirofumi; et al.

Fusion Engineering and Design, 81(8-14), p.1271 - 1276, 2006/02

 Times Cited Count:20 Percentile:80.03(Nuclear Science & Technology)

The design guideline for the blanket is decided to meet the mission of the DEMO plant which is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. To moderate the technological extrapolation, the structural material of reduced activation ferritic steel (F82H), ceramic tritium breeder of Li$$_{2}$$TiO$$_{3}$$ and neutron multiplier of Be are introduced. To improve the economical aspect, the coolant material of the supercritical water with inlet/outlet temperatures of 280/510$$^{circ}$$C, coolant pressure of 25 MPa is chosen. Resultantly the thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirement (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

JAEA Reports

Structural analysis of support structure for ITER vacuum vessel

Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka

JAERI-Tech 2004-068, 27 Pages, 2004/12

JAERI-Tech-2004-068.pdf:7.68MB

ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.

Journal Articles

Design and structural analysis of support structure for ITER vacuum vessel

Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi

Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12

 Times Cited Count:3 Percentile:24.74(Nuclear Science & Technology)

ITER vacuum vessel (VV) is a safety component confining radioactive materials. An independent VV support structure located at the bottom of VV lower port is proposed as an alternative concept, which is deferent from the current reference, i.e., the VV support is directly connected to the toroidal coil (TF coil). This independent concept has two advantages comparing to the reference one: (1) thermal load becomes lower and (2) the TF coil is categorized as a non-safety component. Stress Analyses have been performed to assess the integrity of the VV support structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as an alternative VV support.

Journal Articles

Design and structural analysis of support structure for ITER vacuum vessel

Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi

Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12

The vibration experiments of the support structures with flexible plates for the ITER major components such as toroidal field coil and vacuum vessel were performed using small-sized flexible plates aiming to obtain its basic mechanical characteristics such as dependence of the stiffness on the loading angle. The experimental results obtained by the hammering and frequency sweep tests were agreed each other, so that the experimental method is found to be reliable. In addition, the experimental results were compared with the analytical ones in order to estimate an adequate analytical model. As a result, the bolt connection strongly affected on the stiffness of the support structure. After studies of modeling the bolts, it is found that the analytical results modeling the bolts with finite stiffness only in the axial direction and infinite stiffness in the other directions agree well with the experimental ones. Using this model, the stiffness of the support structure for the ITER major components can be calculated precisely in order to estimate the dynamic behaviors.

Journal Articles

ITER nuclear components, preparing for the construction and R&D results

Ioki, Kimihiro*; Akiba, Masato; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Ezato, Koichiro; Federici, G.*; et al.

Journal of Nuclear Materials, 329-333(1), p.31 - 38, 2004/08

 Times Cited Count:13 Percentile:66.18(Materials Science, Multidisciplinary)

The preparation of the procurement specifications is being progressed for key components. Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 or 30 degree, on flow distribution tests of a two-channel model, on fabrication and testing of FW mockups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

Journal Articles

Examinations on plasma behaviour during disruptions on existing tokamaks and extrapolation to ITER

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Proceedings of 30th EPS Conference on Controlled Fusion and Plasma Physics (CD-ROM), 4 Pages, 2003/07

We examine plasma behaviours during plasma disruptions in detail in JT-60U and other tokamaks to derive appropriate physics guidelines for the behaviours. Their interpretations and their extrapolations to ITER are incorporated into the DINA code, which solves plasma transport and 2D free boundary plasma equilibrium simultaneously with circuit equations for the vacuum vessel and the PF coils. Sensitivity of the plasma behaviours and their impact on the EM force during disruptions due to the range of variation and uncertainty of the experimental data are examined.

Journal Articles

Wave form of current quench during disruptions in Tokamaks

Sugihara, Masayoshi; Lukash, V.*; Kawano, Yasunori; Yoshino, Ryuji; Gribov, Y.*; Khayrutdinov, R.*; Miki, Nobuharu*; Omori, Junji*; Shimada, Michiya

Purazuma, Kaku Yugo Gakkai-Shi, 79(7), p.706 - 712, 2003/07

The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential.

JAEA Reports

Japanese contribution to the design of primary module of shielding blanket in ITER-FEAT

Kuroda, Toshimasa*; Hatano, Toshihisa; Miki, Nobuharu*; Hiroki, Seiji; Enoeda, Mikio; Omori, Junji*; Sato, Shinichi*; Akiba, Masato

JAERI-Tech 2002-098, 136 Pages, 2003/02

JAERI-Tech-2002-098.pdf:24.33MB

no abstracts in English

JAEA Reports

Study for reducing radioactive solid waste at ITER decommissioning period

Sato, Shinichi*; Araki, Masanori; Omori, Junji*; Oono, Isamu*; Sato, Satoshi; Yamauchi, Michinori*; Nishitani, Takeo

JAERI-Tech 2002-083, 126 Pages, 2002/10

JAERI-Tech-2002-083.pdf:4.07MB

It is one of the foremost goals for ITER to demonstrate the attractiveness with regard to safety and environmental potential. This implies that the radioactive materials and waste at decommissioning phase should carefully be treated with prescribed regulations. As possible activities during the Coordinated Technical Activity (CTA), Japanese Participant Team (JA-PT) has proposed a study for searching the possibility of more reduction in the activated level by taking account of minimum material changes while keeping original design concept and structure. Based on it, reassessment of the activation level and the amount of activation volumes will give us positive aspects for public acceptance.

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