S. Jitsukawa

Japan Atomic Energy Agency, Muramatsu, Niigata, Japan

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Publications (147)227.48 Total impact

  • N. Okubo, N. Ishikawa, M. Sataka, S. Jitsukawa
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    ABSTRACT: Microstructure in single crystalline Al2O3 developed during irradiation by swift heavy ions has been investigated. The specimens were irradiated by Xe ions with energies from 70 to 160 MeV at ambient temperature. The fluences were in the range from 1.0 × 1013 to 1.0 × 1015 ions/cm2. After irradiations, X-ray diffractometry (XRD) measurements and cross sectional transmission electron microscope (TEM) observations were conducted. The XRD results indicate that in the initial stage of amorphization in single crystalline Al2O3, high-density Se causes the formation of new planes and disordering. The new distorted lattice planes formed in the early stage of irradiation around the fluence of 5.0 × 1013 ions/cm2 for single crystalline Al2O3 irradiated with 160 MeV-Xe ions. Energy dependence on structural modification was also examined in single crystalline Al2O3 irradiated by swift heavy ions. The XRD results indicate that the swift heavy ion irradiation causes the lattice expansion and the structural modification leading to amorphization progresses above the energy around 100 MeV in this XRD study. The TEM observations demonstrated that amorphization was induced in surface region in single crystalline Al2O3 irradiated by swift heavy ions above the fluence expected from the results of XRD. Obvious boundary was observed in the cross sectional TEM images. The crystal structure of surface region above the boundary was identified to be amorphous and deeper region to be single crystal. The threshold fluence of amorphization was found to be around 1.0 × 1014 ions/cm2 in the case over 80 MeV swift heavy ion irradiation and the fluence did not depend on the crystal structures.
    Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms 11/2013; 314:208-210. DOI:10.1016/j.nimb.2013.05.051 · 1.12 Impact Factor
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    ABSTRACT: Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30 years in China, Europe, India, Japan, Russia and the USA for application in ITER test blanket modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options.
    Journal of Nuclear Materials 11/2013; 442(1-3). DOI:10.1016/j.jnucmat.2012.12.039 · 2.02 Impact Factor
  • Fusion Science and Technology 01/2012; 62(1):139-144. · 0.59 Impact Factor
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    ABSTRACT: Thermal aging properties of reduced activation ferritic/martensitic steel F82H was researched. The aging was performed at temperature ranging from 400°C to 650°C up to 100,000h. Microstructure, precipitates, tensile properties, and Charpy impact properties were carried out on aged materials. Laves phase was found at temperatures between 550 and 650°C and M6C type carbides were found at the temperatures between 500 and 600°C over 10,000h. These precipitates caused degradation in toughness, especially at temperatures ranging from 550°C to 650°C. Tensile properties do not have serious aging effect, except for 650°C, which caused large softening even after 10,000h. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in ductile-to-brittle transition temperature was observed in the 650°C aging. It was caused by the large Laves phase precipitation at grain boundary. Laves precipitates at grain boundary also degrades the upper-shelf energy of the aged materials. These aging test results indicate F82H can be used up to 30,000h at 550°C.
    Fusion Engineering and Design 12/2011; 86(12). DOI:10.1016/j.fusengdes.2011.06.005 · 1.15 Impact Factor
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    ABSTRACT: 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 pathway 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.
    Fusion Engineering and Design 10/2011; 86(6):611-614. DOI:10.1016/j.fusengdes.2011.01.109 · 1.15 Impact Factor
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    ABSTRACT: 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 heat of F82H (Mod3) containing 0.1 % tantalum. The specimens were irradiated up to 20 dpa at 300oC in the 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 optimization 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.
    Journal of Nuclear Materials 10/2011; 417(1-3). DOI:10.1016/j.jnucmat.2011.05.020 · 2.02 Impact Factor
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    ABSTRACT: Embrittlement is known to be caused by P segregation at grain boundaries in Fe alloys. Effects of P substitutions on binding energies and electronic structures of octahedral Fe cluster are investigated using density functional calculations in order to understand the nature of bonding between P and Fe atoms at grain boundaries. The binding energies increase in Fe3P3 and Fe-rich clusters while they decrease in P-rich clusters. The changes in binding energies are closely connected to the charge transfer from Fe to P atoms. The charge transfer leads to both stronger and weaker bonds in mixed Fe–P clusters. The weaker bonds due to less charge cause embrittlement. The calculations indicate that the binding energies and chemical bonding are affected by atomic configurations of P atoms in Fe–P clusters.
    Journal of Nuclear Materials 10/2011; 417(1):1090-1093. DOI:10.1016/j.jnucmat.2010.12.201 · 2.02 Impact Factor
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    ABSTRACT: The status and key issues of reduced activation ferritic/martensitic (RAFM) steels R&D are reviewed as the primary candidate structural material for fusion energy demonstration reactor blankets. This includes manufacturing technology, the as-fabricated and irradiates material database and joining technologies. The review indicated that the manufacturing technology, joining technology and database accumulation including irradiation data are ready for initial design activity, and also identifies various issues that remain to be solved for engineering design activity and qualification of the material for international fusion material irradiation facility (IFMIF) irradiation experiments that will validate the data base.
    Journal of Nuclear Materials 10/2011; 417(1):9-15. DOI:10.1016/j.jnucmat.2011.05.023 · 2.02 Impact Factor
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    ABSTRACT: As a part of the Broader Approach activities, R&D on blanket related materials, reduced-activation ferritic martensitic (RAFM) steels as a structural material, SiCf/SiC composites for flow channel insert in the liquid blanket and/or use as advanced structural material, advanced tritium breeders and neutron multiplier, has been initiated directed at DEMO. As part of the RAFM steel mass production development, a 5 ton heat of RAFM steel (F82H) was procured by Electro Slag Re-melting as the secondary melting method, which was effective in controlling unwanted impurities. An 11 ton heat of EUROFER was also produced. For the SiCf/SiC composite development, NITE- and CVI-SiCf/SiC composites were prepared as reference materials and preliminary mechanical and physical properties were measured. Also compatibility tests between SiC and Pb–17Li have been prepared, related to the He-cooled Li–Pb blanket concept. For the beryllide neutron multiplayer Be–Ti alloy development, large size rods of about 30 mm diameter were fabricated successfully in EU.
    Journal of Nuclear Materials 10/2011; 417(1):1331-1335. DOI:10.1016/j.jnucmat.2010.12.304 · 2.02 Impact Factor
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    ABSTRACT: Reduced activation ferritic/martensitic (RAFM) steels are recognized as the primary candidate structural materials for fusion blanket systems. Because of the possibility of creating sound engineering bases, such as a suitable fabrication technology and a materials database, RAFM steels can be used as structural materials for pressure equipment. Further, the development of an irradiation database in addition to design methodologies for fusion-centered applications is critical when evaluating the applicability of RAFM steels as structural materials for fusion-neutron-irradiated pressure equipment. In the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between the EU and Japan, R&D is underway to optimize RAFM steel fabrication and processing technologies, develop a method for estimating fusion-neutron-irradiation effects, and study the deformation behaviors of irradiated structures. The results of these research activities are expected to form the basis for the DEMO power plant design criteria and licensing. The objective of this paper is to review the BA R&D status of RAFM steel development in Japan, especially F82H (Fe–8Cr–2W–V, Ta). The key technical issues relevant to the design and fabrication of the DEMO blanket and the recent achievements in Japan are introduced.
    Fusion Engineering and Design 10/2011; 86(9):2549-2552. DOI:10.1016/j.fusengdes.2011.04.047 · 1.15 Impact Factor
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    ABSTRACT: The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5 MeV Fe3+ ions up to a dose of 20 dpa at 250 and 380 °C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation.
    Journal of Nuclear Materials 10/2011; 417(s 1–3):270–273. DOI:10.1016/j.jnucmat.2011.01.067 · 2.02 Impact Factor
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    ABSTRACT: The size- and spacing- dependent obstacle strength due to the Cu precipitation in α-Fe is investigated by atomistic simulations, in which the effect on phase transformation of Cu precipitation is considered by a conventional selfguided molecular dynamics (SGMD) method that has an advantage to enhance the conformational sampling efficiency in MD simulations. A sequence of molecular statics simulations of the interaction between a pure edge dislocation and spherical Cu precipitation are performed to investigate the obstacle strength associated with phase transformation. 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.
    Journal of the Society of Materials Science Japan 08/2010; 59(8):583-588. DOI:10.2472/jsms.59.583
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    ABSTRACT: In order to develop a systematic and reasonable concept assuring the structural integrity of components under intense neutron irradiation, two basic tensile properties, true stress–true strain (TS–TS) curves and fracture strain, were investigated on an austenitic stainless steel and martensitic steel. Application of Swift equation is confirmed to a large plastic strain range of TS–TS curves. Fracture strain ɛf data were well correlated as ɛf+ɛ0=const. where ɛ0 is the pre-strain representing the irradiation hardening.Based on those formulations and available experimental information, several critical issues to be dealt with in developing the concept were identified possible reduction in ductility, significant change in mechanical properties, remarkable cyclic softening and other unique cyclic properties observed during a high-cycle fatigue testing, and the redundancy of the plastic collapse concept to bending. Existing structural codes are all based on the assumption that there will be no significant changes in mechanical properties during operation, and of high ductility. Therefore, a new concept for assuring structural integrity is required for application not only to components with high ductility but also components with reduced ductility. First, potential failure modes were identified, and a new and systematic concept was proposed for preventing these modes of failure, introducing a new concept of categorizing the loadings by stability of deformation process to fracture (as type F and M loadings). Based on the basic concept, a detailed concept of how to protect against ductile fracture was given, and loading type-dependent limiting parameters were set.Finally, application of the detailed concept was presented, especially on determination of loading type (in numerical approach, the formulation of TS–TS curves and fracture strain derived above are needed), and on how to determine the limiting parameters as allowable limits. Experiments were done to identify the loading type of a tensile loading acting on a structure with a discontinuity. Tensile loadings acting on an intensely neutron-irradiated flat plate with a hole in the center cause plastic tensile instability and necking at the minimum ligament section but do not initiate any surface crack at the initiation of necking.
    Nuclear Engineering and Design 06/2010; 240(6):1290-1305. DOI:10.1016/j.nucengdes.2010.02.031 · 0.97 Impact Factor
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    ABSTRACT: At temperatures below 400 °C, 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 these changes with neutron dose suggests that some of the reduced activation ferritic/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 property changes is essential to enable these applications. Modelling 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 behaviour after irradiation are discussed. The significance of the models for estimating microstructural change during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.
    Nuclear Fusion 09/2009; 49(11):115006. DOI:10.1088/0029-5515/49/11/115006 · 3.24 Impact Factor
  • Y. Abe, S. Jitsukawa
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    ABSTRACT: 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 α-iron that took place during thermal aging. 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.
    Philosophical Magazine Letters 09/2009; 89(9):535-543. DOI:10.1080/09500830903140735 · 1.27 Impact Factor
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    ABSTRACT: The effects of He on the fracture behavior of reduced-activation ferritic/martensitic steels, including oxide dispersion-strengthened (ODS) steels and F82H, was determined by characterizing the microstructural evolution in and fracture behavior of these steels after He implantation up to 1000appm at around 550°C. He implantation was carried out by a cyclotron with a beam of 50MeV α-particles. In the case of F82H, the ductile-to-brittle transition temperature (DBTT) increase induced by He implantation was about 70°C and the grain boundary fracture surface was only observed in the He-implanted area of all the ruptured specimens in brittle manner. By contrast, no DBTT shift or fracture mode change was observed in He-implanted 9Cr-ODS and 14Cr-ODS steels. Microstructural characterization suggested that the difference in the bubble formation behavior of F82H and ODS steels might be attributed to the grain boundary rupture of He-implanted F82H.
    Journal of Nuclear Materials 04/2009; 386:241-244. DOI:10.1016/j.jnucmat.2008.12.102 · 2.02 Impact Factor
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    ABSTRACT: Breeding blankets are the most important components in DEMO. The DEMO blanket has to withstand high neutron flux typically 15–30 dpa/year under continuous operation. Therefore integrated and effective development of blanket structural materials and breeding/multiplying materials is essential in the blanket development for DEMO. In parallel to the ITER program, broader approach (BA) activities are initiated by EU and Japan. Based on the common interest of each party towards DEMO, R&D on reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiCf/SiC composites which have potential for use in DEMO blankets, advanced tritium breeders and neutron multiplier for DEMO blankets, and tritium technologies including tritium behavior studies in advanced materials for DEMO blanket applications will be carried out as a part of the BA activities.
    Journal of Nuclear Materials 04/2009; 386:405-410. DOI:10.1016/j.jnucmat.2008.12.146 · 2.02 Impact Factor
  • Yukio Miwa, Shiro Jitsukawa, Takashi Tsukada
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    ABSTRACT: 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 temperatures in oxygenated or hydrogenated water. Test specimens of F82H were heat-treated at various temperature conditions, or were cold-worked to simulate radiation hardening and machined to make single edge notch, or were neutron-irradiated at 493K to 3.4dpa. 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 573K in oxygenated water. In irradiated specimen, TGSCC occurred, when SSRT was conducted at 573K in hydrogenated (DH=1ppm) water or when the notched specimen was tested by SSRT at 573K in oxygenated (DO=10ppm) water.
    Journal of Nuclear Materials 04/2009; 386:703-707. DOI:10.1016/j.jnucmat.2008.12.332 · 2.02 Impact Factor
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    ABSTRACT: The irradiation behavior of Li2TiO3 under a fusion reactor environment was simulated by simultaneous irradiation of Li2TiO3 by the triple ion beams and the respective single ion beams of O2+, He+ and H+. The microstructural changes in Li2TiO3 caused by the irradiation were measured by FT-IR photoacoustic spectroscopy. The results suggest that the amount of TiO2 formed is proportional to the dpa and that the method of irradiation does not affect the dependence of formation of TiO2. On the other hand, the amount of defects and/or radiolytic products generated by irradiation, which is considered to trap hydrogen near the surface, is found to be affected by the method of irradiation. Such phenomena are believed to affect the tritium release behavior from Li2TiO3, and durability of Li2TiO3 and compatibility of Li2TiO3 with other components of the breeder blanket such as structural materials in the fusion reactor system under operation.
    Journal of Nuclear Materials 04/2009; 386:1065-1067. DOI:10.1016/j.jnucmat.2008.12.279 · 2.02 Impact Factor
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    ABSTRACT: Significant progress has been achieved in the international research effort on reduced activation ferritic/martensitic steels for fusion structural applications. Because this class of steels is the leading structural material for test blankets in ITER and future fusion power systems, the range of ongoing research activities is extremely broad. Since, it is not possible to discuss all relevant work in this brief review, the objective of this paper is to highlight significant issues that have received recent attention. These include: (1) efforts to measure and understand radiation-induced hardening and embrittlement at temperatures ⩽400 °C, (2) experiments and modeling to characterize the effects of He on microstructural evolution and mechanical properties, (3) exploration of approaches for increasing the high-temperature (>550 °C) creep resistance by introduction of a high-density of nanometer scale dispersoids or precipitates in the microstructure, (4) progress toward structural design criteria to account for loading conditions involving both creep and fatigue, and (5) development of nondestructive examination methods for flaw detection and evaluation.
    Journal of Nuclear Materials 04/2009; 386:411-417. DOI:10.1016/j.jnucmat.2008.12.323 · 2.02 Impact Factor

Publication Stats

2k Citations
227.48 Total Impact Points

Institutions

  • 1994–2013
    • Japan Atomic Energy Agency
      • Nuclear Science and Engineering Directorate
      Muramatsu, Niigata, Japan
  • 2001–2002
    • Kyoto University
      • Institute of Advanced Energy
      Kioto, Kyōto, Japan
  • 1991–1992
    • Oak Ridge National Laboratory
      • Materials Science and Technology Division
      Oak Ridge, Florida, United States