Okubo, Nariaki; Ishikawa, Norito; Sataka, Masao; Jitsukawa, Shiro
Nuclear Instruments and Methods in Physics Research B, 314, p.208 - 210, 2013/11
Microstructure in single crystal AlO specimens developed during irradiation by 70-160 MeV-Xe ions has been examined with transmission electron microscope (TEM). Amorphization was observed around 800 nm depth through from ion-beam incident surface of the specimen. The amorphization was also evaluated with XRD technique. TEM observation also indicated that tracks were formed by irradiation at the deeper region than that of the amorphous layers. The number density and size of the tracks have been decreased with the depth from the ion-incident surface. It suggests that amorphization has been occurred by the overlapping of the tracks. It was also obvious that the thickness of the amorphous layers were smaller than those observed in poly-crystal alumina specimens. This may be interpreted by the difference of electronic energy loss in the specimen; lower energy loss for single crystal specimens for their crystallographic orientation against to the incident ion-beam.
Abe, Yosuke; Tsuru, Tomohito; Jitsukawa, Shiro
Materials Research Society Symposium Proceedings, Vol.1535 (Internet), 7 Pages, 2013/04
In this study, in addition to adequately model the 1D reaction kinetics of SIA loops in the framework of a production bias model, reaction kinetics associated with carbon impurity atoms present in -iron have been formulated to take into account the trapping effect of glissile SIA loops by vacancy-carbon (V-C) complexes that have been shown to have strong bindings with SIA loops by atomistic simulations. Results of calculated defect accumulation behavior of neutron irradiated -iron show that the developed CD model can successfully reproduce the number densities of SIA loops and vacancy clusters when the applied impurity concentration is the same order as experimental one. This indicates that the assumed mechanism for the trapping of glissile SIA loops by V-C complexes is reasonable. The dependences of irradiation dose, dose rate, and temperature are discussed in detail.
Abe, Yosuke; Jitsukawa, Shiro; Okubo, Nariaki; Matsui, Hideki*; Tsukada, Takashi
Effects of Radiation on Nuclear Materials; 25th Volume (ASTM STP 1547), p.313 - 337, 2013/01
It is known that the mechanical properties degradation of reactor pressure vessel steels caused by neutron irradiation is partly due to the formation of nanometer-size solute and point-defect (PD) clusters. Therefore, the rationalization of radiation induced effects on the microstructure and their consequences on the material properties by developing predictive models is thus of great importance. Cluster dynamics (CD) simulation with rate equations has been used to estimate the long-term evolution of point defect clusters. We have extended the CD simulation code to allow hopping motion for all the self-interstitial atom (SIA) clusters to be mobile. Results of calculation on 316 stainless steel and -iron have been compared. The difference and underlying mechanism of microstructural evolution between austenitic stainless steel and -iron is discussed with regard to the dose dependence.
Abe, Yosuke; Suzudo, Tomoaki; Jitsukawa, Shiro; Tsuru, Tomohito; Tsukada, Takashi
Fusion Science and Technology, 62(1), p.139 - 144, 2012/07
Static and dynamic interactions between SIA loops and carbon impurities as a major impurity in -iron was studied using atomic-scale computer simulations to clarify possible mechanisms of trapping of SIA loops by carbon impurities. It was found that bindings of carbon-vacancy complexes with SIA loops are so strong when they locate inside the loop habit plane. This situation would be realized below 450 K at which carbon-vacancy complexes become to dissociate. At higher temperatures, it can be expected that the dissociated interstitial carbon atoms tend to migrate towards the loop periphery due to the internal stress field associated with SIA loops, and this could attribute to the continuous 1D motion of SIA loops with dragging impurities as observed by experiments. A long-term microstructural evolution induced by radiation damage was also simulated by taking into account the trapping and detrapping rates of SIA loops deduced from the knowledge obtained in this study.
Shiba, Kiyoyuki; Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro
Fusion Engineering and Design, 86(12), p.2895 - 2899, 2011/12
Aging properties of reduced activation ferritic/martensitic steel F82H was researched at temperature ranging from 400C to 650C up to 100,000 hr. Microstructure, tensile, and Charpy properties were carried out. Laves was found at temperatures between 550 and 650C and MC carbides were found at the temperatures between 500 and 600C over 10,000 hr. These precipitates caused degradation in toughness, especially at temperatures ranging from 550C to 650C. Tensile properties do not have serious aging effect, except for 650C, which caused large softening even after 10.000 hr. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in DBTT caused by the large Laves phase precipitation at grain boundary was observed in the 650C aging. Laves precipitates at grain boundary also degrades the USE of the aged materials. These aging test results provide F82H can be used up to 30,000 hr at 550C.
Garin, P.*; Diegele, E.*; Heidinger, R.*; Ibarra, A.*; Jitsukawa, Shiro; Kimura, Haruyuki; Mslang, A.*; Muroga, Takeo*; Nishitani, Takeo; Poitevin, Y.*; et al.
Fusion Engineering and Design, 86(6-8), p.611 - 614, 2011/10
This paper summarizes the proposals and findings of the IFMIF Specification Working Group established to update the Users requirements and top level specifications for the Facility. Special attention is given to the different roadmaps of fusion path way towards power plants, of materials R&D and of facilities and their interactions. The materials development and validation activities on structural materials, blanket functional materials and non-metallic materials are analyzed and specific objectives and requirements to be implemented in IFMIF are proposed. Emphasis is made in additional potential validation activities that can be developed in IFMIF for ITER TBM qualification as well as for DEMO-oriented mock-up testing.
Nishitani, Takeo; Tanigawa, Hiroyasu; Nozawa, Takashi; Jitsukawa, Shiro; Nakamichi, Masaru; Hoshino, Tsuyoshi; Yamanishi, Toshihiko; Baluc, N.*; Mslang, A.*; Lindou, R.*; et al.
Journal of Nuclear Materials, 417(1-3), p.1331 - 1335, 2011/10
As a part of the Broader Approach (BA) activities, the research and development on blanket related materials and tritium technology have been initiated toward DEMO by Japan and EU. Recently, those five R&D items have progressed substantially in Japan and EU. As a preparatory work aiming at the RAFM steel muss-production development, a 5-ton heat of RAFM steel (F82H) was procured with the Electro Slag Re-melting as a secondary melting. The result of the double notch tensile test method for the NITE-SiC/SiC specimen indicated notch insensitivity and very minor size effect on proportional limit tensile stress and fracture strength. For the fabrication technology development of beryllide neutron multiplayer pebbles, Be- Ti inter-metallic pebbles have been sintered directly from the mixed powder of Be and Ti in Japan.
Nakazawa, Tetsuya; Igarashi, Takahiro; Tsuru, Tomohito; Kaji, Yoshiyuki; Jitsukawa, Shiro
Journal of Nuclear Materials, 417(1-3), p.1090 - 1093, 2011/10
It is well known that impurities in iron which segregate to grain boundaries can dramatically change physical and chemical properties. Phosphorous, which is one of impurities, segregates at grain boundaries under thermal or irradiation environments, and brings about the intergranual embrittlement. In this study, influences of phosphorus substitutions for binding energies and electronic structures of octahedral iron cluster are investigated computationally using density functional calculations in order to understand the nature of bonding between phosphorus and iron at grain boundaries. The values of binding energies of cluster are increasing with the phosphorous substitutions. The increases are due to Fe-P bond strengthened by the charge transfer from Fe atom to P atom. On the other hand the calculated bond orders give that Fe-Fe bonds are weakened. Thus, the embrittlement induced with the segregation of P due to irradiation is considered to be associated with weakened Fe-Fe bonds.
Okubo, Nariaki; Sokolov, M. A.*; Tanigawa, Hiroyasu; Hirose, Takanori; Jitsukawa, Shiro; Sawai, Tomotsugu; Odette, G. R.*; Stoller, R. E.*
Journal of Nuclear Materials, 417(1-3), p.112 - 114, 2011/10
Irradiation hardening and fracture toughness of reduced-activation ferritic/martensitic steel F82H after irradiation were investigated with a focus on changing the fracture toughness transition temperature as a result of several heat treatments. The specimens were standard F82H-IEA (IEA), F82H-IEA with several heat treatments (Mod1 series) and a higher tantalum containing (0.1%) heat of F82H (Mod3). The specimens were irradiated up to 18 dpa at 300 C in High Flux Isotope Reactor under a collaborative research program between JAEA/US-DOE. The results of hardness tests showed that irradiation hardening of IEA was comparable with that of Mod3. However, the fracture toughness transition temperature of Mod3 was lower than that of IEA. The transition temperature of Mod1 was also lower than that of the IEA heat. These results suggest that tightening of specifications on the heat treatment condition and modification of the minor alloying elements seem to be effective to reduce the fracture toughness transition temperature after irradiation.
Tanigawa, Hiroyasu; Shiba, Kiyoyuki; Mslang, A.*; Stoller, R. E.*; Lindau, R.*; Sokolov, M. A.*; Odette, G. R.*; Kurtz, R. J.*; Jitsukawa, Shiro
Journal of Nuclear Materials, 417(1-3), p.9 - 15, 2011/10
ITER construction was started, and R&D toward DEMO shifted to more practical stage. On this stage, the candidate material for DEMO blanket have to be the one which have sound engineering bases to be ready for engineering designing activity for DEMO reactor in 10 years. Reduced activation ferritic/martensitic (RAFM) steels, such as F82H (Fe-8Cr-2W-0.2V-0.04Ta) or EUROFER97 (Fe-9Cr-1W-0.2V- 0.12Ta), is the only material which currently have enough potential to meet this requirement, and selected as the target material in the R&D on materials engineering for DEMO blanket under the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between EU and Japan. In this paper, current status of RAFM R&D is overviewed especially on fabrication technology, inspection/testing technology, and material database. Overview on irradiation effect study is also provided.
Tsuru, Tomohito; Abe, Yosuke; Kaji, Yoshiyuki; Tsukada, Takashi; Jitsukawa, Shiro
Zairyo, 59(8), p.583 - 588, 2010/08
The size- and spacing- dependent obstacle strength due to the Cu precipitation in -Fe is investigated by atomistic simulations, in which the effect on phase transformation of Cu precipitation is considered by a conventional self-guided molecular dynamics (SGMD) method that has an advantage to enhance the conformational sampling efficiency in MD simulations. It was shown that the SGMD method can accelerate calculating the bcc to 9R structure transformation of a small precipitate, enabling the transformation without introducing any excess vacancies. Such metallographic structures increase the obstacle strength through strong pinning effects as a result of the complicated atomic rearrangement within the Cu precipitation.
Suzuki, Kazuhiko; Jitsukawa, Shiro; Okubo, Nariaki; Takada, Fumiki
Nuclear Engineering and Design, 240(6), p.1290 - 1305, 2010/06
Jitsukawa, Shiro; Suzuki, Kazuhiko; Okubo, Nariaki; Ando, Masami; Shiba, Kiyoyuki
Nuclear Fusion, 49(11), p.115006_1 - 115006_8, 2009/11
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.
Abe, Yosuke; Jitsukawa, Shiro
Philosophical Magazine Letters, 89(9), p.535 - 543, 2009/09
The self-guided molecular dynamics (SGMD) method which can enhance the conformational sampling efficiency in MD simulations, was applied in investigating the phase transformation of Cu precipitate in -iron that took place during thermal ageing. It was shown that the SGMD method can accelerate calculating the bcc to 9R structure transformation of a small precipitate (even 4.0 nm in size), enabling the transformation without introducing any excess vacancies. The size dependence of the transformation also agreed with that seen in previous experimental studies.
Nishitani, Takeo; Tanigawa, Hiroyasu; Jitsukawa, Shiro; Nozawa, Takashi; Hayashi, Kimio; Yamanishi, Toshihiko; Tsuchiya, Kunihiko; Mslang, A.*; Baluc, N.*; Pizzuto, A.*; et al.
Journal of Nuclear Materials, 386-388, p.405 - 410, 2009/04
The establishment of the breeding blanket technology is one of the most important engineering issues on the DEMO development. For the DEMO blanket, developments of the structural materials and functional materials such as tritium breeder and neutron multiplier. Which should be used under the savior circumstance such as high neutron fluence, high temperature and strong magnetic field, are urgent issues. In the Broader Approach activities initiated by EU and Japan, developments of reduced activation ferritic martensitic steels as a DEMO blanket structural material, SiC/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, are planed as common interest issues of EU and Japan. This paper describes the overview of the development program.
Miwa, Yukio; Jitsukawa, Shiro; Tsukada, Takashi
Journal of Nuclear Materials, 386-388, p.703 - 707, 2009/04
In order to examine the stress corrosion cracking (SCC) susceptibility of reduced activation ferritic/martensitic steel, F82H, slow strain rate test (SSRT) was performed at various temperature in oxygenated or hydrogenated water. Test specimens of F82H were heat-treated at various temperature conditions, or were cold-worked to imitate radiation hardening and machined to make single edge notch, or were neutron-irradiated at 493 K to 3.4 dpa. It was found that in unirradiated specimen, IGSCC occurred when specimen was normalized only, and TGSCC occurred when cold-worked (over 23%) and notched specimen was tested by SSRT at 573 K in oxygenated water. In irradiated specimen, TGSCC occurred, when SSRT was conducted at 573 K in hydrogenated (DH = 1 ppm) water or when the notched specimen was tested by SSRT at 573 K in oxygenated (DO = 10 ppm) water.
Abe, Yosuke; Jitsukawa, Shiro
Philosophical Magazine, 89(4), p.375 - 388, 2009/02
A combination of simulated annealing with Langevin molecular dynamics and basin-hopping with occasional jumping (BHOJ) technique was used to systematically determine the most stable configurations of self-interstitial atom (SIA) clusters ( 1-38) in -iron. In addition to the original BHOJ technique, we introduced an additional long jumping process in which a randomly selected less-bounded atom is moved to a neighboring site of another SIA in the cluster to enhance the probability of locating the global minimum structure. With the obtained putative lowest energy structures, the binding energies as a function of cluster size were estimated. We also determined the sizes of particular stable clusters based on their geometrical symmetry. Furthermore, the values were extrapolated based on accurately determined formation energies, and are available for immediate use in kinetic Monte Carlo or rate theory models.
Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Cevolani, S.*; Chida, Teruo*; Ciotti, M.*; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; et al.
Fusion Engineering and Design, 83(7-9), p.1007 - 1014, 2008/12
This paper describes the latest design of liquid lithium target system in IFMIF. Design requirement of the Li target is to provide a stable Li jet with a speed of 20 m/s to handle an averaged heat flux of 1 GW/m. A double reducer nozzle and a concaved flow are applied to the target design. On Li purification, a cold trap and two kinds of hot trap are applied to control impurities below permissible levels. Nitrogen concentration shall be controlled below 10 wppm by one of the hot trap. Tritium concentration shall be controlled below 1 wppm by an yttrium hot trap. To maintain reliable continuous operation, various diagnostics are attached to the target assembly. Among the target assembly, a back-plate made of RAFM is located in the most severe region of neutron irradiation (50 dpa/y). Therefore, two design options of replaceable back wall and their remote handling systems are under investigation.
Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki; Kasada, Ryuta*; Wakai, Eiichi; Serizawa, Hisashi*; Kawahito, Yosuke*; Jitsukawa, Shiro; Kimura, Akihiko*; Kono, Yutaka*; et al.
Fusion Engineering and Design, 83(10-12), p.1471 - 1476, 2008/12
Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. F82H, which were developed and studied in Japan, was designed with an emphasis on high temperature properties and weldability. The database on F82H properties is currently the most extensive available among the existing RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER Test Blanket Module (TBM) suggested by recent achievements in Japan.
Ushigusa, Kenkichi; Seki, Masahiro; Ninomiya, Hiromasa; Norimatsu, Takayoshi*; Kamada, Yutaka; Mori, Masahiro; Okuno, Kiyoshi; Shibanuma, Kiyoshi; Inoue, Takashi; Sakamoto, Keishi; et al.
Genshiryoku Handobukku, p.906 - 1029, 2007/11
no abstracts in English