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Journal Articles

Conceptual design for Japan Sodium-cooled Fast Reactor, 4; Developmental study of steel plate reinforced concrete containment vessel for JSFR

Negishi, Kazuo; Hosoya, Takusaburo; Sato, Kenichiro*; Somaki, Takahiro*; Matsuo, Ippei*; Shimizu, Katsusuke*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9418_1 - 9418_7, 2009/05

An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results.

JAEA Reports

Conceptual design and related R&D on ITER mechanical based primary pumping system

Tanzawa, Sadamitsu; Hiroki, Seiji; Abe, Tetsuya; Shimizu, Katsusuke*; Inoue, Masahiko*; Watanabe, Mitsunori*; Iguchi, Masashi*; Sugimoto, Tomoko*; Inohara, Takashi*; Nakamura, Junichi*

JAEA-Technology 2008-076, 99 Pages, 2008/12

JAEA-Technology-2008-076.pdf:35.19MB

The primary vacuum pumping system of the International Thermonuclear Experimental Reactor (ITER) exhausts a helium (He) ash resulting from the DT-burn with excess DT fueling gas, as well as performing a variety of functions such as pump-down, leak testing and wall conditioning. A mechanical based vacuum pumping system has some merits of a continuous pumping, a much lower tritium inventory, a lower operational cost and easy maintenance, comparing with a cryopump system, although demerits of an indispensable magnetic shield and insufficient performance for hydrogen (H$$_{2}$$) pumping are well recognized. To overcome the demerits, we newly fabricated and tested a helical grooved pump (HGP) unit suitable for H$$_{2}$$ pumping at the ITER divertor pressure of 0.1-10 Pa. Through this R&D, we successfully established many design and manufacturing databases of large HGP units for the lightweight gas pumping. Based on the databases, we conceptually designed the ITER vacuum pumping system mainly comprising the HGP with an optimal pump unit layout and a magnetic shield. We also designed conceptually the reduced cost (RC)-ITER pumping system, where a compound molecular pump combining turbine bladed rotors and helical grooved ones was mainly used. The ITER mechanical based primary pumping system proposed has eventually been a back-up solution, whereas a cryopump based one was formally selected to the ITER for construction.

Journal Articles

Basic analysis of weldability and machinability of structural materials for ITER toroidal field coils

Onozuka, Masanori*; Shimizu, Katsusuke*; Urata, Kazuhiro*; Kimura, Masahiro*; Kadowaki, Hirokazu*; Okamoto, Mamoru*; Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi

Fusion Engineering and Design, 82(5-14), p.1431 - 1436, 2007/10

 Times Cited Count:2 Percentile:18.75(Nuclear Science & Technology)

no abstracts in English

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:37.26(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

Thermo-structural analysis and design consideration of the replaceable backwall in IFMIF liquid lithium target

Nakamura, Hiroo; Ida, Mizuho; Chida, Teruo; Shiba, Kiyoyuki; Shimizu, Katsusuke*; Sugimoto, Masayoshi

Journal of Nuclear Materials, 367-370(2), p.1543 - 1548, 2007/08

 Times Cited Count:2 Percentile:18.75(Materials Science, Multidisciplinary)

The IFMIF is an accelerator-based intense neutron source for testing candidate materials of fusion reactor. Intense neutrons are emitted inside the Li flow through a backwall. The backwall made of 316L stainless steel or RAFM is attached to the target assembly with a lip seal welded by YAG laser. Since the backwall is operating under a severe neutron irradiation of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm$$^{3}$$, thermo-structural design is one of critical issues in a target design. Thermal stress was calculated by ABAQUS code. As a permissible stress, yield strength at 300$$^{circ}$$C was used. In a case of the 316 stainless steel backwall, a maximum thermal stress was more than the permissible stress(164MPa). On the other hand, in case of the F82H backwall, a maximum thermal stress is was below the permissible stress(455MPa). Therefore, F82H is recommended as a backwall material.

JAEA Reports

Thermo-structural analysis of backwall in IFMIF lithium target

Nakamura, Hiroo; Ida, Mizuho; Shimizu, Katsusuke*; Sugimoto, Masayoshi

JAEA-Technology 2007-008, 28 Pages, 2007/03

JAEA-Technology-2007-008.pdf:4.05MB

This report describes results of thermo-structural analysis of a backwall in IFMIF lithium target performed during FY 2003-2006. The IFMIF is an accelerator-based intense neutron source for testing candidate materials for fusion reactors. Intense neutrons are emitted inside the Li flow through a backwall. The backwall is made of 316L stainless steel or RAFM. Since the backwall is operating under a severe neutron irradiation of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm$$^{3}$$, thermo-structural design is one of critical issues in a target design. Thermal stress was calculated using the ABAQUS code. In a case of the 316L stainless steel backwall, the maximum thermal stress was beyond the permissible stress. On the other hand, in a case of the F82H backwall, a maximum thermal stress was 289 MPa below the permissible stress (455 MPa). Therefore, F82H is recommended as the backwall material.

JAEA Reports

Study on assembly techniques and procedures for ITER tokamak device

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi*; Ue, Koichi*; Shimizu, Katsusuke*; Onozuka, Masanori*

JAEA-Technology 2006-034, 85 Pages, 2006/06

JAEA-Technology-2006-034.pdf:9.18MB

The International Thermonuclear Experimental Reactor (ITER) tokamak is composed of many kinds of components. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of these components are required to be a high accuracy of $$pm$$3mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements. Based on the above background, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology concept based on the available knowledge and experiences of the installation of the large components, in order to improve the IT (International Team) design toward the more realistic one. As a result, we show the realistic tokamak assembly procedures and techniques to be able to assemble the large and heavy ITER tokamak with high accuracy. (1)Assembly and alignment of the toroidal field coil with high accuracy. (2)Simplification of the assembly procedures, and the jigs/tools and procedures to reduce the misalignment. (3)Assembly procedures and techniques for the vacuum vessel to reduce the weld distortion. (4)Supporting procedures and techniques of the vacuum vessel sector to prevent the toridal field coil from the deformation due to the dead weight of the vacuum vessel sector. (5)Datum points and lines for the required positions and assembly tolerances during tokamak assembly.

Journal Articles

Thermal and thermal-stress analyses of IFMIF liquid lithium target assembly

Ida, Mizuho*; Nakamura, Hiroo; Shimizu, Katsusuke*; Yamamura, Toshio*

Fusion Engineering and Design, 75-79, p.847 - 851, 2005/11

 Times Cited Count:3 Percentile:24.27(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Review of JAERI activities on the IFMIF liquid lithium target in FY2004

Nakamura, Hiroo; Ida, Mizuho*; Matsuhiro, Kenjiro; Fischer, U.*; Hayashi, Takumi; Mori, Seiji*; Nakamura, Hirofumi; Nishitani, Takeo; Shimizu, Katsusuke*; Simakov, S.*; et al.

JAERI-Review 2005-005, 40 Pages, 2005/03

JAERI-Review-2005-005.pdf:3.52MB

The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based Deuterium-Lithium (Li) neutron source to produce intense high energy neutrons (2 MW/m$$^{2}$$) up to 200 dpa and a sufficient irradiation volume (500 cm$$^{3}$$) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, radioactive species such as 7Be, tritium and activated corrosion products are generated. In addition, back wall operates under severe conditions of neutron irradiation damage (about 50 dpa/y). In this paper, the thermal and thermal stress analyses, the accessibility evaluation of the IFMIF Li loop, and the tritium inventory and permeation of the IFMIF Li loop are summarized as JAERI activities on the IFMIF target system performed in FY2004.

Journal Articles

Present status of the liquid lithium target facility in the international fusion materials irradiation facility (IFMIF)

Nakamura, Hiroo; Riccardi, B.*; Loginov, N.*; Ara, Kuniaki*; Burgazzi, L.*; Cevolani, S.*; Dell'Ocro, G.*; Fazio, C.*; Giusti, D.*; Horiike, Hiroshi*; et al.

Journal of Nuclear Materials, 329-333(1), p.202 - 207, 2004/08

 Times Cited Count:14 Percentile:66.17(Materials Science, Multidisciplinary)

International Fusion Materials Irradiation Facility (IFMIF), being developed by EU, JA, RF and US, is a deuteron-lithium (Li) reaction neutron source for fusion materials testing. In the end of 2002, 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors has been completed. This paper describes these KEP tasks results. To evaluate Li flow characteristics, a water and Li flow experiments have been done. To develop Li purification system, evaluation of nitrogen and tritium gettering materials have been done. Conceptual design of remote handling and basic experiment have been donde. In addition, safety analysis and diganostics design have been done. In the presentation, the latest design and future prospects will be also summarized.

Journal Articles

Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

Onozuka, Masanori*; Takeda, Nobukazu; Nakahira, Masataka; Shimizu, Katsusuke*; Nakamura, Tomomichi*

Fusion Engineering and Design, 69(1-4), p.757 - 762, 2003/09

 Times Cited Count:2 Percentile:19.02(Nuclear Science & Technology)

The dynamic behavior of the ITER tokamak assembly has been investigated. Three experimental models have been considered to validate the numerical analysis methods for the dynamic events, mainly seismic events. A 1/8-scaled tokamak model, which is based on the 1998 ITER design, is under construction. Non-linear vibration characteristics, such as damping, can only be identified by a full-scale model. Therefore, a full-scale gravity support structure for the coil system has been designed and will be tested. In addition, for the sub-scaled tokamak model, the VV is assumed to be a rigid structure. This assumption is to be verified using a 1/20-scaled model. The above experimental models and their testing conditions have analytically and numerically evaluated. For example, both the static and dynamic spring constants obtained by static analysis and eigen-value analysis, respectively, were evaluated to be in good agreement.

Journal Articles

Fabrication of simulated leak test apparatus for developing water leak detection method applicable to in-vessel water cooling channels in fusion machines

Hiroki, Seiji; Tanzawa, Sadamitsu; Arai, Takashi; Abe, Tetsuya; Shimizu, Katsusuke*; Nakatani, Junnosuke*; Kuribayashi, Shizuma*

Shinku, 44(3), P. 329, 2001/03

no abstracts in English

JAEA Reports

Geometrical influence on repeated impact durability of alumina insulation film coated by plasma spraying method

Kanari, Moriyasu*; Abe, Tetsuya; Tanzawa, Sadamitsu; Shimizu, Katsusuke*; *; *

JAERI-Research 99-012, 21 Pages, 1999/02

JAERI-Research-99-012.pdf:2.26MB

no abstracts in English

JAEA Reports

Durability of alumina electrical insulation films by plasma spraying coating method under repeated impact loads

Kanari, Moriyasu*; Abe, Tetsuya; Enoeda, Mikio; *; *; Shimizu, Katsusuke*; *; Takatsu, Hideyuki

JAERI-Research 98-029, 23 Pages, 1998/06

JAERI-Research-98-029.pdf:2.51MB

no abstracts in English

Journal Articles

Design of the ITER vacuum vessel

Ioki, Kimihiro*; G.Johnson*; Shimizu, Katsusuke*; D.Williamson*

Fusion Engineering and Design, 27, p.39 - 51, 1995/00

 Times Cited Count:13 Percentile:76.26(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Design and development of supporting system for ITER/CDA blanket box structure

Nishio, Satoshi; *; Shimizu, Katsusuke*; Koizumi, Koichi; Abe, Tetsuya; Tada, Eisuke

JAERI-M 93-091, 92 Pages, 1993/03

JAERI-M-93-091.pdf:2.69MB

no abstracts in English

JAEA Reports

Aseismatic analysis of ITER vacuum vessel, 1; Modeling and eigenvalue analysis

Futakawa, Masatoshi; Koizumi, Koichi; Shimizu, Katsusuke*; Takatsu, Hideyuki; ; *

JAERI-M 92-164, 58 Pages, 1992/11

JAERI-M-92-164.pdf:1.92MB

no abstracts in English

JAEA Reports

Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor(ITER)

Shimizu, Katsusuke*; *; Koizumi, Koichi; *; Nishio, Satoshi; Sasaki, Takashi*; Tada, Eisuke

JAERI-M 92-135, 139 Pages, 1992/09

JAERI-M-92-135.pdf:3.74MB

no abstracts in English

JAEA Reports

Design of divertor support structure

*; Nishio, Satoshi; *; Koizumi, Koichi; Shimizu, Katsusuke*; Sasaki, Takashi*; Tada, Eisuke

JAERI-M 92-108, 62 Pages, 1992/07

JAERI-M-92-108.pdf:1.59MB

no abstracts in English

JAEA Reports

Structural design of cryostat and port penetration of International Thermonuclear Experimental Reactor(ITER)

*; *; Shimizu, Katsusuke*; *; Okawa, Yoshinao; Shibanuma, Kiyoshi; Tada, Eisuke

JAERI-M 92-094, 102 Pages, 1992/07

JAERI-M-92-094.pdf:2.3MB

no abstracts in English

28 (Records 1-20 displayed on this page)