Nonlinear Fracture Mechanics Analysis of Surface Crack Propagation in Fusion Reactor First Wall by Cyclic High Heat Flux Irradiation

Professor Hiroshi Okada and his team from the Department of Mechanical Engineering, Faculty of Science and Technology, Tokyo University of Science, Japan, are engaged in the field of computational fracture mechanics. This is an area of computational engineering that refers to the creation of numerical methods to approximate the crack evolutions predicted by new classes of fracture mechanics models. For many years, it has been used to determine stress intensity factors and, more recently, has expanded into the simulation of crack nucleation and propagation. In their work, the researchers are proposing new methods for fracture mechanics analysis and solid mechanics analysis.

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Plasma-Sprayed Beryllium on Macro-Roughened Substrates for Fusion Reactor High Heat Flux Applications

Journal of Thermal Spray Technology ◽

10.1007/s11666-006-9011-6 ◽

2007 ◽

Vol 16 (1) ◽

pp. 96-103 ◽

Cited By ~ 3

Author(s):

Kendall J. Hollis ◽

Brian D. Bartram ◽

Manfred Roedig ◽

Dennis Youchison ◽

Richard Nygren

Keyword(s):

Heat Flux ◽

Fusion Reactor ◽

High Heat ◽

High Heat Flux ◽

Plasma Sprayed

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Post-mortem analysis of HIP bonded first wall panel made of SS316 and DS-Cu after high heat flux testing

Fusion Technology 1996 ◽

10.1016/b978-0-444-82762-3.50098-7 ◽

1997 ◽

pp. 511-514 ◽

Cited By ~ 1

Author(s):

Toshihisa Hatano ◽

Kiyoshi Fukaya ◽

Masayuki Dairaku ◽

Toshimasa Kuroda ◽

Hideyuki Takatsu

Keyword(s):

Heat Flux ◽

High Heat ◽

High Heat Flux ◽

Post Mortem ◽

First Wall ◽

Wall Panel

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Fabrication and high heat flux testing of plasma sprayed beryllium ITER first wall mock-ups

Journal of Nuclear Materials ◽

10.1016/s0022-3115(98)00078-6 ◽

1998 ◽

Vol 258-263 ◽

pp. 252-257 ◽

Cited By ~ 9

Author(s):

R.G Castro ◽

K.E Elliot ◽

R.D Watson ◽

D.L Youchison ◽

K.T Slattery

Keyword(s):

Heat Flux ◽

High Heat ◽

High Heat Flux ◽

First Wall ◽

Plasma Sprayed

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Evaluation of Strength on Dissimilar Metal Joints for ITER First Wall Components

ASME 2010 Pressure Vessels and Piping Conference: Volume 1 ◽

10.1115/pvp2010-26008 ◽

2010 ◽

Author(s):

H. Nishi ◽

M. Enoeda ◽

T. Hirose ◽

D. Tsuru ◽

H. Tanigawa

Keyword(s):

Heat Flux ◽

Stainless Steel ◽

Stress Singularity ◽

Dissimilar Materials ◽

High Heat ◽

High Heat Flux ◽

First Wall ◽

Cu Alloy ◽

Quarter Plane ◽

Plane Joint

The first wall (FW) of ITER blanket includes beryllium (Be) armor tiles joined to CuCrZr heat sink with stainless steel cooling tube and backing plate in order to improve plasma performance and reduce thermal stress. Therefore dissimilar materials joints are indispensable for fabricating the high heat flux components. Since these joints must withstand thermal and mechanical loads caused by the plasma and electromagnetic force, it is important to evaluate the strength and thermal fatigue life of dissimilar materials joints. When the dissimilar materials joints are subjected by external force and thermal loading, the stress of the joint may indicate singularity at the interface edge. Since the stress singularity may lower the strength of joints, the singularity is evaluated numerically for the various materials combinations and joint configuration to be used in high heat flux components of fusion reactors in this investigation. Moreover, tensile test and elasto-plastic FEM analysis are performed to investigate the fracture behavior of Be/Cu alloy and stainless steel/ Cu alloy obtained the FW mock-up. The results reveal two singular solutions of type rpj−1 for a half-plane bonded to a quarter-plane joint and the singularity is larger than that of a bonded quarter-planes joint. From the viewpoint of stress singularity, the configuration of bonded quarter-planes joint is better than the half-plane bonded to a quarter-plane joint. The singularity for W/Cu alloy combination is large compared to other combination of materials. Especially the singularity of stainless steel/ Cu alloy is very small. Tensile specimen of Be/CuCrZr joint fractured at the bonding interface due to the stress singularity. For the stainless steel/ Cu alloy, however, the specimens fractured at the Cu alloy region apart from the interface.

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High heat flux testing of first wall mock-ups with and without neutron irradiation

Nuclear Materials and Energy ◽

10.1016/j.nme.2016.02.003 ◽

2016 ◽

Vol 9 ◽

pp. 41-45 ◽

Cited By ~ 5

Author(s):

G. Pintsuk ◽

B. Bellin ◽

A. Gervash ◽

J. Linke ◽

N. Litunovsky ◽

...

Keyword(s):

Heat Flux ◽

Neutron Irradiation ◽

High Heat ◽

High Heat Flux ◽

First Wall

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Overview of the R&D Activities of Water Cooled Ceramic Breeder Blanket

18th International Conference on Nuclear Engineering: Volume 6 ◽

10.1115/icone18-30143 ◽

2010 ◽

Author(s):

Mikio Enoeda ◽

Takanori Hirose ◽

Hisashi Tanigawa ◽

Daigo Tsuru ◽

Akira Yoshikawa ◽

...

Keyword(s):

Heat Flux ◽

Martensitic Steel ◽

Side Wall ◽

High Heat ◽

Hydraulic Performance ◽

High Heat Flux ◽

Fabrication Technology ◽

First Wall ◽

Key Technologies ◽

Real Scale

This paper overviews the R&D activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing R&D of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m2, 80 cycles with the coolant condition as DEMO, 15 MPa 300 °C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been progress toward the establishment of fabrication technology of WCCB TBM.

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