Anisotropic creep property related to non-spherical shape of mechanically alloyed powder of oxide dispersion strengthened F82H

Sakasegawa, Hideo; Nakajima, Motoki*; Kato, Taichiro*; Nozawa, Takashi*; Ando, Masami*

Materials Today Communications (Internet), 40, p.109659_1 - 109659_8, 2024/08

https://doi.org/10.1016/j.mtcomm.2024.109659

Nanometric oxide particles play an important role in improving the creep property of Oxide Dispersion Strengthened (ODS) steels. In our previous research, we examined a microstructural feature known as prior particle boundary (PPB). PPB refers to the surface of mechanically alloyed (MA) powders before consolidation. We revealed that the ODS steel with fine PPBs produced from smaller MA powders, exhibited shorter creep rupture times, compared to that with coarse PPBs produced from larger MA powders. The size of MA powders had an impact on the creep property. In this study, we examined the shape of MA powders, which were non-spherical shapes. Such shapes have the potential to induce anisotropic creep behavior. We conducted small punch creep tests on specimens with two different orientations to study the possible anisotropy. The results revealed that the creep rupture times varied depending on the orientation of specimen, thus indicating anisotropic creep property.

Recent improvement and evaluation of radiation resistance and magnetic properties of high entropy alloys and their applications

Wakai, Eiichi; Noto, Hiroyuki*; Shibayama, Tamaki*; Furuya, Kazuyuki*; Wakui, Takashi; Ando, Masami*; Makimura, Shunsuke*; Ishida, Taku*

Science Talks (Internet), 8, p.100278_1 - 100278_4, 2023/12

https://doi.org/10.1016/j.sctalk.2023.100278

High entropy alloys tend to combine high strength with good ductility due to their inherent properties. This material is considered as a promising new material not only for higher-performance future general industrial applications, but also for increasing the durability and range of application of radiation-affected equipment in nuclear and radiation environments, and has been rapidly gaining attention in recent years. In this study, two types of high-entropy alloys (Fe-Mn-V-Cr-Al-C and Fe-Si-W-Cr-V) composed of low-radioactive elements (without Ni and Co) were prepared and their basic properties were evaluated for application as new functional materials to be used under radiation in high-energy accelerator target system components, nuclear reactors, fusion reactors, etc. and their basic properties were evaluated. The two materials under development in this study have unique properties in the following respects. The former is expected to be developed as a basic research for high-power motor materials as a new structural material and magnetic properties sharing the features of high strength and low radiation. On the other hand, the latter is expected to be applied as a new functional material in new engineering fields by mixing tungsten, which has the highest melting point among metallic elements, with vanadium, which has a considerably higher melting point, to raise the melting point of the alloy and to design an alloy with high strength.

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.

Effects of heat treatments and addition of minor elements of boron and nitrogen on mechanical properties and microstructures of reduced-activation ferritic/martensitic steel

Wakai, Eiichi; Ando, Masami; Okubo, Nariaki

Journal of Plasma and Fusion Research SERIES, Vol.11, p.104 - 112, 2015/03

http://www.jspf.or.jp/JPFRS/PDF/Vol11/jpfrs2015_11-104.pdf

The reduced-activation ferritic/martensitic (RAFM) steels for the fusion DEMO reactor have been developing from around the 1980s. RAFM steels are the first candidate materials for the first wall and blanket structure of fusion DEMO reactors, the target back-plate and the target assembly of IFMIF. In this study, two subjects had been examined and are summarized as below: (1) Effect of initial heat treatment on the microstructures and mechanical properties of RAFM steels, including irradiation damage, is very important to design the fusion DEMO reactors and also control the changes of mechanical properties after the irradiation. (2) Effects of He and H production on the microstructures and mechanical properties of RAFM steels, including irradiation damage, are essential in the evaluation of design of fusion DEMO reactor, and we have to check and evaluate them in Fusion irradiation environment like IFMIF.

Radiation-induced effects in physical properties of materials, 2-6; Evaluation of irradiation creep for F82H steel by using pressurized tubes

Ando, Masami; Nozawa, Takashi; Hirose, Takanori; Tanigawa, Hiroyasu

Purazuma, Kaku Yugo Gakkai-Shi, 90(1), p.64 - 67, 2014/01

http://ci.nii.ac.jp/naid/110009688791

Reduced activation ferritic/martensitic steel (RAFM) is a candidate for the material of DEMO blanket structure. The irradiation creep behavior of F82H and JLF-1 steel has been measured at 300, 400 and 500C up to 5 dpa using helium-pressurized creep tubes irradiated in HFIR. These tubes were pressurized with helium to hoop stress levels of 0400 MPa at the irradiation temperature. The results for F82H and JLF-1 with a 400 MPa hoop stress detected small creep strains ( 0.25%) after irradiation at 300C. Irradiation creep rate (creep strain/dose) was tendency to be a similar behavior for high-dose irradiated RAFM specimens in FFTF. In this paper, a procedure of irradiation creep test & evaluation was also summarized.

Irradiation hardening in F82H irradiated at 573 K in the HFIR

Hirose, Takanori; Okubo, Nariaki; Tanigawa, Hiroyasu; Ando, Masami; Sokolov, M. A.*; Stoller, R. E.*; Odette, G. R.*

Journal of Nuclear Materials, 417(1-3), p.108 - 111, 2011/10

https://doi.org/10.1016/j.jnucmat.2010.12.044

This paper summarizes recent results of the irradiation experiments focused on F82H and its modified steels irradiated at 573 K. The materials used in this research were F82H-IEA and its modified steels. Post irradiation mechanical tests revealed that irradiation hardening of F82H is saturated by 9 dpa and the as-irradiated proof stress is less than 1 GPa. The deterioration of total elongation was also saturated by 9 dpa. Irradiation response of F82H-mod3, which is stable to temperature instability during material production and HIP treatment, was very similar to that of F82H-IEA, and negative impacts of extra tantalum was not observed. Therefore it can be an attractive option for the structural materials for blanket components manufactured by HIP.

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Summary report of joint research in Japan on BA DEMO R&D activity 2008 (Joint research)

Tanigawa, Hiroyasu; Nozawa, Takashi; Ando, Masami

JAEA-Review 2009-053, 89 Pages, 2010/03

JAEA-Review-2009-053.pdf:22.39MB

R&D on DEMO engineering was started from 2007 as the main activity of a DEMO design and R&D coordination center activity which was one of the international nuclear fusion energy research center (IFERC) activity of "Broader Approach" (BA) activity. In the DEMO engineering R&D activity, five "generic" R&D items were selected, which attract both Japan and EU interest and not specific to a DEMO design, i.e., (1) R&D on SiC/SiC Composites: (2) R&D on Tritium Technology: (3) R&D on Materials Engineering for DEMO Blanket: (R&D on reduced activation ferritic/martensitic steel: RAFM) (4) R&D on Advanced Neutron Multiplier for DEMO Blanket: (5) R&D on Advanced Tritium Breeders for DEMO Blanket. In first phase (2007 -2009), some part of preliminary R&D, which was defined in procurement arrangement (PA) between Japan and EU, is conducted by JAEA university collaboration. This report summarized the results from joint research activities about R&D on SiC/SiC composite and RAFM in 2008.

Irradiation effects on reduced activation ferritic/martensitic steels; Tensile, impact, fatigue properties and modelling

Jitsukawa, Shiro; Suzuki, Kazuhiko; Okubo, Nariaki; Ando, Masami; Shiba, Kiyoyuki

Nuclear Fusion, 49(11), p.115006_1 - 115006_8, 2009/11

https://doi.org/10.1088/0029-5515/49/11/115006

Irradiation often causes hardening and reduction of elongation as well as toughness degradation to a considerable degree. Data, however, indicate that these changes remain in manageable ranges for ITER-TBM application. Moreover, the saturation tendency of the changes with neutron dose suggests that some of the reduced activation martensitic steels are feasible even for future DEMO applications. It is also stressed that the development of a design methodology that is compatible with the large irradiation induced changes is essential to enable these applications. Modeling activities for the macroscopic mechanical response are expected to play key roles in design methodology development. Macroscopic models of plasticity (a constitutive equation) and cyclic softening behavior after irradiation are discussed. Significance of models to estimate microstructural changes during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

https://doi.org/10.1088/0029-5515/49/7/075029

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).