Microstructural evolution in tungsten binary alloys under proton and self-ion irradiations at 800C

Miyazawa, Takeshi; Kikuchi, Yuta*; Ando, Masami*; Yu, J.-H.*; Yabuuchi, Kiyohiro*; Nozawa, Takashi*; Tanigawa, Hiroyasu*; Nogami, Shuhei*; Hasegawa, Akira*

Journal of Nuclear Materials, 575, p.154239_1 - 154239_11, 2023/03

https://doi.org/10.1016/j.jnucmat.2023.154239

Development of benchmark reduced activation ferritic/martensitic steels for fusion energy applications

Tanigawa, Hiroyasu; Gaganidze, E.*; Hirose, Takanori; Ando, Masami; Zinkle, S. J.*; Lindau, R.*; Diegele, E.*

Nuclear Fusion, 57(9), p.092004_1 - 092004_13, 2017/06

https://doi.org/10.1088/1741-4326/57/9/092004

The current status of RAFM developments and evaluations, including the applicability of joining technologies, is reviewed. The technical challenges and potential risks of utilizing RAFM steels as the structural material of in-vessel components are discussed, and possible mitigation methodology is introduced. The discussion suggests that deuterium-tritium fusion neutron irradiation effects currently need to be treated as an ambiguity factor which could be incorporated within the safety factor. The safety factor will be defined by the engineering design criteria which are not yet developed with regard to irradiation effects and some high temperature process, and the operating time condition of the in-vessel component will be defined by the condition at which those ambiguities due to neutron irradiation become too large to be acceptable, or by the critical condition at which 14 MeV fusion neutron irradiation effects is expected to become different from fission neutron irradiation effects.

Evaluation of impacts of stress triaxiality on plastic deformability of RAFM steel using various types of tensile specimen

Kato, Taichiro; Ohata, Mitsuru*; Nogami, Shuhei*; Tanigawa, Hiroyasu

Fusion Engineering and Design, 109-111(Part B), p.1631 - 1636, 2016/11

https://doi.org/10.1016/j.fusengdes.2015.11.003

Plastic deformability of material shows general tendency to decrease due to become hard and brittle. Also, the plastic deformability tends to decrease as the stress triaxiality of a parameter to evaluate the magnitude of plastic constraint increases. Therefore, it is necessary to accurately understand the ductility loss limit of RAFM in order to conduct the structural design assessment of a fusion reactor demo blanket. In this study, plastic deformability of RAFM was evaluated the impacts of stress triaxiality on variation of tensile specimen shape and testing conditions. In the results, the fracture was defined as not the point of macro-crack but that of micro-crack. It was confirmed that the true strain rate significantly increases in the vicinity of the point of micro-crack. The relationships between the fracture ductile and stress triaxiality of the full size tensile specimen and the miniature size one were shown on the single curve regardless of the specimen size.

Hydrogen behavior in primary precipitate of F82H steel; Atomistic calculation based on the density functional theory

Watanabe, Yoshiyuki; Iwakiri, Hirotomo*; Murayoshi, Norihiko*; Kato, Daiji*; Tanigawa, Hiroyasu

Plasma and Fusion Research (Internet), 10, p.1205086_1 - 1205086_2, 2015/12

https://doi.org/10.1585/pfr.10.1205086

In this paper, formation energy of isolated hydrogen atom in CrC has been theoretically investigated with atomistic calculation based on the density functional theory. The lowest calculated formation energy of a hydrogen atom is -0.48 eV with a trigonal bipyramidal configuration surrounded by five regular Cr lattice atoms. A comparison of the formation energy of hydrogen atom in bcc-iron may indicate that hydrogen atoms in F82H steel are more energetically favorable in CrC-based precipitate rather than Fe-based matrix, leading to an increase of the tritium retention in the precipitate.

Mechanical properties of TIG and EB weld joints of F82H

Hirose, Takanori; Sakasegawa, Hideo; Nakajima, Motoki; Tanigawa, Hiroyasu

Fusion Engineering and Design, 98-99, p.1982 - 1985, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.06.133

As a R&D activity on materials engineering for DEMO blanket in ITER-BA activity, characterization of F82H weld joints prepared with Tungsten-Inert-Gas (TIG) and electron beam (EB) have been investigated. In this work, 50 mm thick plates of F82H were welded using both processes. A similar-metal was employed as a filler for TIG welding. Post-weld-heat-treatment was conducted according to the conditions for Grade 91 defined as ASME P-No.15E, Group No.1. Although the maximum and the minimum hardness of the both joint are similar, the hardness distribution is quite different. The width of EB welds were smaller than that of TIG, and the hardness of EB weld metal was 10% higher than that of TIG. In the TIG welds, the strongest part was heat affected zone (HAZ) heated above phase transformation temperature, Ac1 and the hardness was very similar to the weld metal of EB joint, 280Hv. The hardness of TIG weld metal was around 260 Hv. Both welds demonstrated the smallest hardness, 180 Hv in the HAZ heated below Ac1 temperature. As a investigation of manufacturing process of box fabrication, second EB weld bead was perpendicularly put on the first EB bead. As a result, the second weld did not weaken the HAZ, but reduced the hardness of the weld metal to 260 Hv.

Modification of vacuum plasma sprayed tungsten coating on reduced activation ferritic/martensitic steels by friction stir processing

Tanigawa, Hiroyasu; Ozawa, Kazumi; Morisada, Yoshiaki*; Noh, S.*; Fujii, Hidetoshi*

Fusion Engineering and Design, 98-99, p.2080 - 2084, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.04.059

The vacuum plasma spray (VPS) technique has been investigated as the most practical method to form Tungsten (W) layer as a plasma facing material in fusion devices. The issues are the thermal conductivity and the strength of VPS-W, i.e., the thermal conductivity of VPS-W were significantly lower than that of the bulk W, and the hardness of VPS-W is much less than that of the bulk W. These are mainly caused by the porous structure of VPS-W. In order to solve these issues, friction stir processing (FPS) was applied on VPS-W in this study. It was suggested that FSP can contribute to significant improvement both in mechanical and thermal properties of VPS-W coating.

Neutronics analysis for fusion DEMO reactor design

Someya, Yoji; Tobita, Kenji; Tanigawa, Hisashi; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in past several years, which includes new findings regarding design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of LiTiO and BeTi pebbles. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR.

Stress envelope of silicon carbide composites at elevated temperatures

Nozawa, Takashi; Kim, S.*; Ozawa, Kazumi; Tanigawa, Hiroyasu

Fusion Engineering and Design, 89(7-8), p.1723 - 1727, 2014/10

https://doi.org/10.1016/j.fusengdes.2013.12.032

A SiC/SiC composite is a promising candidate material for the advanced fusion DEMO blanket. For the design of the DEMO, the stability of high-temperature strength of SiC/SiC composites needs to be identified. Additionally, strength anisotropy needs to be clarified because of its unique fabric architecture. This study therefore aims to evaluate mechanical properties by various modes at elevated temperatures, eventually providing a stress envelope for the design. A P/W Tyranno-SA3 fiber reinforced CVI SiC matrix composite with multilayered SiC/PyC interface was evaluated in this study. Tensile and compressive tests were conducted by the SSTT specifically arranged for the high-temperature use. In-plane shear properties were contrarily estimated by the off-axial tensile method assuming that the mixed mode failure criterion is valid for composites. All tests were performed in vacuum. The preliminary test results indicate no degradation of both proportional limit stress (PLS) and the ultimate tensile strength at temperatures below 1000C. Similarly, no significant degradation of high-temperature compressive and in-plane shear properties were identified, finally providing the stress envelope at elevated temperatures for the design.

Compatibility of Ni and F82H with liquid Pb-Li under rotating flow

Kanai, Akihiko*; Park, C.*; Noborio, Kazuyuki*; Kasada, Ryuta*; Konishi, Satoshi*; Hirose, Takanori; Nozawa, Takashi; Tanigawa, Hiroyasu

Fusion Engineering and Design, 89(7-8), p.1653 - 1657, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.03.008

Study of safety features and accident scenarios in a fusion DEMO reactor

Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.

Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.04.062

After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.