Numerical evaluation of crack propagation of ITER first wall with an initial interfacial defect

Suzuki, Satoshi; Enoeda, Mikio; Matsuda, Hirokazu*; Hiramatsu, Hideki*; Kuroda, Toshimasa*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(3), p.365 - 369, 2007/09

https://doi.org/10.3327/taesj.J06.060

The first wall of ITER will be fabricated by means of HIP method for the bonding of cooling tubes and a copper alloy heat sink. A ultrasonic testing (UT) is adopted as a non-destructive inspection method for the bonding interface as one of acceptance tests of the first wall components. Therefore, clarification of defect size criteria is one of critical issues for the soundness of the first wall. Thermo-mechanical behavior of an initial defect at the bonded interface of the first wall was numerically analyzed. J-integral was calculated to evaluate the propagation behavior of the interfacial defects under thermal loading. As a result, it was found that the initial defect size of 10mm 20mm in semi-elliptic shape was unlikely to propagate. This defect size is more than ten times larger than a detection limit of present UT techniques, and it can be resulted that the UT method presently available is sufficient to detect such harmful initial defects of the ITER first wall.

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

https://doi.org/10.1016/j.fusengdes.2005.08.100

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 LiTiO and neutron multiplier of Be are introduced. To improve the economical aspect, the coolant material of the supercritical water with inlet/outlet temperatures of 280/510C, 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.

Overview on materials R&D activities in Japan towards ITER construction and operation

Takatsu, Hideyuki; Sato, Kazuyoshi; Hamada, Kazuya; Nakahira, Masataka; Suzuki, Satoshi; Nakajima, Hideo; Kuroda, Toshimasa*; Nishitani, Takeo; Shikama, Tatsuo*; Shu, Wataru

Journal of Nuclear Materials, 329-333(1), p.178 - 182, 2004/08

https://doi.org/10.1016/j.jnucmat.2004.04.340

This paper presents an overview on ITER-supporting materials research and development activities and major achievements in Japan during the period from the Co-ordinated Technical Activities to date. In view of the completed engineering design of ITER during the Engineering Design Activities period, research and development efforts since then have been focused: those for reduction of component fabrication cost; those in support of domestic preparations of a structural technical code for construction; those necessary for operation, and been extended to component-level testing rather than pure material testing. They cover materials Research and Development for in-vessel components, vacuum vessel, cryogenic steels of superconducting mgnet and diagnostics components. Major achievements in each research and development area are highlighted and their impact or implication to the design, construction and operation of ITER is presented.

Test blanket program in ITER

Kuroda, Toshimasa*; Enoeda, Mikio; Akiba, Masato

FAPIG, (164), p.17 - 23, 2003/07

no abstracts in English

Design improvements and R&D achievements for vacuum vessel and in-vessel components towards ITER construction

Ioki, Kimihiro*; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Gervash, A.*; Ibbott, C.*; Jones, L.*; et al.

Nuclear Fusion, 43(4), p.268 - 273, 2003/04

https://doi.org/10.1088/0029-5515/43/4/308

Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same, there have been several detailed design improvements resulting from efforts to raise reliability, to improve maintainability and to save money. One of the most important achievements in the VV R&D has been demonstration of the necessary fabrication and assembly tolerances. Recently the deformation due to cutting of the port extension was measured and it was shown that the deformation is small and acceptable. Further development of advanced methods of cutting, welding and NDT on a thick plate have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R&D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated.

Development of Be/DSCu HIP bonding and thermo-mechanical evaluation

Hatano, Toshihisa; Kuroda, Toshimasa*; Barabash, V.*; Enoeda, Mikio

Journal of Nuclear Materials, 307-311(2), p.1537 - 1541, 2002/12

https://doi.org/10.1016/S0022-3115(02)01071-1

no abstracts in English

Development of HIP technique for bonding of CuCrZr with stainless steel and beryllium for application to the ITER first wall

Hatano, Toshihisa; Enoeda, Mikio; Kuroda, Toshimasa*; Akiba, Masato

JAERI-Tech 2002-075, 59 Pages, 2002/10

JAERI-Tech-2002-075.pdf:17.38MB

no abstracts in English

Thermal cycle test of elemental mockups of ITER breeding blanket

Yanagi, Yoshihiko*; Kosaku, Yasuo; Hatano, Toshihisa; Kuroda, Toshimasa*; Enoeda, Mikio; Akiba, Masato

JAERI-Tech 2002-046, 45 Pages, 2002/05

JAERI-Tech-2002-046.pdf:2.61MB

no abstracts in English

Conceptual design of solid breeder blanket system cooled by supercritical water

Enoeda, Mikio; Ohara, Yoshihiro; Akiba, Masato; Sato, Satoshi; Hatano, Toshihisa; Kosaku, Yasuo; Kuroda, Toshimasa*; Kikuchi, Shigeto*; Yanagi, Yoshihiko*; Konishi, Satoshi; et al.

JAERI-Tech 2001-078, 120 Pages, 2001/12

JAERI-Tech-2001-078.pdf:8.3MB

This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. This conceptual design study was performed to determine the updated strategy and goal of the R&D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology.

Safety analysis of ITER test blanket module for water cooled blanket with pebble bed breeder

Enoeda, Mikio; Kuroda, Toshimasa*; Moriyama, Koichi*; Ohara, Yoshihiro

Journal of Nuclear Science and Technology, 38(11), p.921 - 929, 2001/11

https://doi.org/10.3327/jnst.38.921

Test module testing in ITER is one of the most important mile-stone for development of the DEMO blanket. In the design of test modules in ITER, it is very important to show that test modules do not cause additional safety concern to ITER. This work has been performed for the evaluation of the substantial safety of Test Module of Water Cooled Solid Blanket, which is the current candidate blanket for the DEMO blanket in Japan. Major issues of the evaluation were establishment of post accident cooling in TM, hydrogen gas generation by Be-steam reaction, and pressure increase and spilled water amount by Loss of Coolant Accident (LOCA) event. The evaluation was performed to derive the upper bound of consequences in significant events, of which scenario can be assumed by the similarity of the safety analysis of Shielding Blanket.