S. Jitsukawa

Japan Atomic Energy Agency, Muramatsu, Niigata, Japan

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Publications (161)219.32 Total impact

  • S. Suzuki · S. Sato · M. Suzuki · H. Kinoshita · S. Jitsukawa · H. Tanigawa
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    ABSTRACT: Small specimen test technology (SSTT) is a key technique in evaluating the irradiation performance of reduced-activation ferritic/martensitic (RAF/M) steels used in fusion demonstration plants. Because SSTT results are rather sensitive to surface finishing conditions (mechanical damage, roughness, and dimensional accuracy), the effect of surface finishing on tensile data was examined. Sheet tensile specimens were 5 mm long, 1.5 mm wide, and 0.76 mm thick gage sections (SS-J3 type) prepared from 15-mm-thick plates of RAF/M steel F82H (F82H-B07 heat). SS-J3 specimens were obtained from the plates by wire electro-discharge machining (WEDM). The surfaces of the specimens were finished using several techniques: polishing with abrasive paper, chemical polishing, and polishing with alumina suspensions. With these techniques, four types of surface-finished specimens were prepared: (1) WEDM specimens, (2) specimens finished by abrasive paper polishing, (3) specimens finished by abrasive paper polishing followed by alumina suspension polishing, and (4) specimens finished by abrasive paper polishing followed by chemical and alumina suspension polishing. Tensile tests were conducted at a nominal strain rate of 3.33 × 10-4/s. The effects of surface finishing were not strong. Specimens finished by method 4 exhibited the highest yield stress level. Those finished by method 3 exhibited similar but slightly smaller strength values. The tensile results were fairly uniform along the thickness of the plates. The ultimate tensile strength values obtained from the specimens finished using method 3 were 645 MPa (standard deviation of 16 MPa) for the 15-mm-thick plates. The area reduction of the 15-mm-thick plates was 0.76 in natural strain. Copyright © 2015 by ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959.
    No preview · Article · Jan 2015
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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.
    No preview · Article · Nov 2013 · Journal of Nuclear Materials
  • 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.
    No preview · Article · Nov 2013 · Nuclear Instruments and Methods in Physics Research Section B Beam Interactions with Materials and Atoms
  • Yosuke Abe · Tomohito Tsuru · Shiro Jitsukawa
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    ABSTRACT: Cluster dynamics (CD) modeling has been used to estimate the long-term evolution of point defect (PD) clusters. However, previous studies have often simplified the governing equations by assuming the maximum size of mobile self-interstitial atom (SIA) clusters and by ignoring the one-dimensional (1D) reaction kinetics of SIA loops. They have also conducted parameter fittings, such as the clustered fraction and the maximum size of clusters produced by collision cascade, to reproduce experimental data. In this study, in addition to modeling the 1D motion of SIA loops in the framework of the production bias model (PBM), reaction rates associated with carbon impurity atoms present in alpha iron were formulated to consider the trapping effect of one-dimensionally migrating SIA loops by a vacancy-carbon (V-C) complex that was shown to have strong bindings with SIA loops by previous atomistic simulations. Calculations results for neutron-irradiated alpha iron showed that the developed CD model can successfully reproduce the saturation trend of the number density of immobile SIA loops in contrast to the prediction using a model without the trapping effect.
    No preview · Article · Jan 2013 · MRS Online Proceeding Library
  • Shiro Jitsukawa · Yosuke Abe · Kazuhiko Suzuki · Nariaki Okubo
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    ABSTRACT: Neutron irradiation often introduces severe changes in the mechanical properties of austenitic and martensitic steels (e.g., reduction of elongation below 400°C). This affects the mechanical responses of the reactor components. A large amount of hot cell work, however, is required in order to examine the mechanical response of intensely irradiated components experimentally, as well as to obtain materials irradiation data for the estimation of the component behavior. The development of a methodology with which to estimate the mechanical response of such components based on knowledge of the irradiation-induced microstructural changes and models of the post-irradiation mechanical properties is therefore an effective way to evaluate the structural integrity of an intensely irradiated structural component with minimal effort. A methodology for simulating the microstructural changes of face-centered cubic metals during irradiation using molecular dynamics and rate equation (RE) calculations has been developed. For RE calculation, the capture radius of the point defect clusters has been obtained through in situ ion irradiation experiments. The flow stress level is estimated from the dispersed barrier hardening equation with calculated microstructural data. By means of correlating the limited data from irradiation experiments with the calculated results (relation between the flow stress level and the microstructure of a heat), the flow stress levels of the heat can be estimated accurately as functions of the damage level and temperature. A Swift type constitutive equation with the concept of an equivalent plastic strain of irradiation hardening is proposed from the work hardening behavior of irradiated steels. By using this equation, it is possible to estimate the deformation and ductile fracture conditions of intensely irradiated components. This is a preliminary multi-scale method for estimating the mechanical response of irradiated components.
    No preview · Article · Jan 2013
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    ABSTRACT: It is known that the degradation of mechanical properties of reactor pressure vessel steels caused by neutron irradiation is partly due to the formation of nanometer-size solute and point-defect (PD) clusters. Depending on temperature, PDs produced by collision cascades under neutron irradiation can migrate and either recombine or agglomerate to form larger defect clusters, greatly affecting the microstructure evolution and thus the mechanical properties of the material. Therefore, studying the rationalization of radiation-induced effects on the microstructure and their influence on the material properties through the development of predictive models is of great importance. A cluster dynamics (CD) simulation based on rate equations has been used to estimate the long-term evolution of point-defect clusters, i.e., clusters of self-interstitial atoms (SIAs) and those of vacancies and precipitations containing solute atoms. We have extended a CD simulation code to account for the possibility of all SIA clusters migrating three dimensionally, to reproduce the agglomeration of point-defects to form clusters during irradiation with collision cascades in austenitic stainless steel. We have also performed a parametric study of a production bias model, which can take into account the effects of fast one-dimensional motion of SIA loops, of defect accumulation processes in neutron-irradiated α-iron. It is found that formulations that take into account proper reaction kinetics for different materials can successfully reproduce the microstructure evolution under neutron irradiation.
    No preview · Article · Jan 2013
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    ABSTRACT: It is known that the presence of even a small amount of impurity in interstitial positions can, depending on temperature, have a drastic influence on the onedimensional (1-D) motion of self-interstitial atom (SIA) loops, and thus, on the accumulation of radiation damage in materials. In this study, atomic-scale computer simulations based on a recently developed optimization technique have been performed to evaluate the binding energies of SIA loops with interstitial carbon, a vacancycarbon (V-C) complex, and a vacancy as a function of loop size in a-iron. While weak and strong attractive interactions are found when an interstitial carbon atom and a vacancy, respectively, are located on the perimeter of an SIA loop, the interactions for both quickly weaken approaching the loop center. In contrast, for a wide range of loop sizes, significantly higher binding energies are obtained between an SIA loop and a V-C complex located within the habit plane of the loop. A cluster dynamics model was developed by taking into account the trapping effects of V-C complexes on 1-D migrating SIA loops, and preliminary calculations were performed to demonstrate the validity of the assumed trapping mechanism through a comparison of the microstructural evolution with experimental data in neutronirradiated a-iron.
    No preview · Article · Jul 2012 · Fusion Science and Technology
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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.
    No preview · Article · Dec 2011 · Fusion Engineering and Design
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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.
    No preview · Article · Oct 2011 · Fusion Engineering and Design
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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.
    No preview · Article · Oct 2011 · Journal of Nuclear Materials
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    T. Nakazawa · T. Igarashi · T. Tsuru · Y. Kaji · S. Jitsukawa
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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.
    Full-text · Article · Oct 2011 · Journal of Nuclear Materials
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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.
    No preview · Article · Oct 2011 · Journal of Nuclear Materials
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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.
    No preview · Article · Oct 2011 · Journal of Nuclear Materials
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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.
    No preview · Article · Oct 2011 · Fusion Engineering and Design
  • C. Liu · C. Yu · N. Hashimoto · S. Ohnuki · M. Ando · K. Shiba · S. Jitsukawa
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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.
    No preview · Article · Oct 2011 · Journal of Nuclear Materials
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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.
    No preview · Article · Aug 2010 · Journal of the Society of Materials Science Japan
  • Kazuhiko Suzuki · Shiro Jitsukawa · Nariaki Okubo · Fumiki Takada
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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.
    No preview · Article · Jun 2010 · Nuclear Engineering and Design
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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.
    No preview · Article · Sep 2009 · Nuclear Fusion
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    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.
    Full-text · Article · Sep 2009 · Philosophical Magazine Letters
  • A. Hasegawa · M. Ejiri · S. Nogami · M. Ishiga · R. Kasada · A. Kimura · K. Abe · S. Jitsukawa
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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.
    No preview · Article · Apr 2009 · Journal of Nuclear Materials