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Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi
JAEA-Testing 2023-004, 76 Pages, 2024/03
JAEA-Testing-2023-004.pdf:2.08MB
This manual describes the methods for conducting material tests in air, argon gas, and sodium, and for organizing the data obtained, as a part of the development of high-temperature structural design technology for fast reactors. This manual reflects the revision of test methods in Japanese Industrial Standards (JIS) to the "FBR Metallic Materials Test Manual, PNC TN241 77-03" published in 1977 and the "FBR Metallic Materials Test Manual (Revised Edition), JNC TN9520 2001-001" published in 2001. Also, it was written with reference to the recommended room temperature / elevated temperature tensile test method by the Japan Society of Mechanical Engineers (JSME) and the test standard for the elevated-temperature low-cycle fatigue test method by the Society of Materials Science, Japan (JSMS), which are the standard for material test methods in the domestic academic society.
C in Mod.9Cr-1Mo steel to 1
10
cyclesToyota, Kodai; Imagawa, Yuya; Onizawa, Takashi; Kato, Shoichi; Furuya, Yoshiyuki*
Nihon Kikai Gakkai Rombunshu (Internet), 89(928), p.23-00206_1 - 23-00206_15, 2023/12
In order to design fast reactors, it is necessary to consider high cycle fatigue of structural materials up to 110
cycles; to evaluate high cycle fatigue at 110 cycles, it is necessary to develop a best-fit fatigue curve applicable up to 110 cycles. In this study, high cycle fatigue tests were conducted under strain-controlled conditions and ultrasonic fatigue tests were also conducted to develop a high cycle fatigue evaluation method for Mod.9Cr-1Mo steel, which is a candidate material for fast reactor structural materials. Based on the test results, the best-fit fatigue curves were extended and the applicability of the JSME best-fit fatigue curves up to 110 cycles was verified.
Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 60(3), p.288 - 298, 2023/03
Times Cited Count：4 Percentile：78.52(Nuclear Science & Technology)JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.
Yano, Yasuhide; Hashidate, Ryuta; Tanno, Takashi; Imagawa, Yuya; Kato, Shoichi; Onizawa, Takashi; Ito, Chikara; Uwaba, Tomoyuki; Otsuka, Satoshi; Kaito, Takeji
JAEA-Data/Code 2021-015, 64 Pages, 2022/01
JAEA-Data-Code-2021-015.pdf:2.6MB
From a view point of practical application of fast breeder reactor cycles, which takes advantage of safety and economic efficiency and makes a contribution of volume reduction and mitigation of degree of harmfulness of high-level radioactive waste, it is necessary to develop fuel cladding materials for fast reactors (FRs) in order to achieve high-burnup. Oxide dispersion strengthened (ODS) steel have been studied for use as potential fuel cladding materials in FRs owing to their excellent resistance to swelling and their high-temperature strength in Japan Atomic Energy Agency. It is very important to establish the materials strength standard in order to apply ODS steels as a fuel cladding. Therefore, it is necessary to acquire the mechanical properties such as tensile, creep rupture strength tests and so on. In this study, tensile and creep rupture strengths of 9Cr-ODS steel claddings were evaluated using by acquired these data. Because of the phase transformation temperature of 9Cr-ODS steel, temperature range for the evaluation was divided into two ones at AC1 transformation temperature of 850C.
Kutsuna, Hideki; Koda, Yuya; Imagawa, Yasuhiro
JAEA-Review 2014-045, 38 Pages, 2015/01
JAEA-Review-2014-045.pdf:67.33MB
Fugen Decommissioning Engineering Center, in planning and carrying out our decommissioning technical development, has been establishing "Technical special committee on Fugen decommissioning" which consists of the members well-informed, aiming to make good use of Fugen as a place for technological development which is opened inside and outside the country, as the central point in the energy research and development base making project of Fukui prefecture, and to utilize the outcome in our decommissioning to the technical development effectively. This report compiles presentation materials "The Current Situation of Fugen Decommissioning", "Evaluation for Dismantlement Work Experience on Turbine System and its Future Plan" and "Completion of Taking-Out Heavy Water from Fugen", presented in the 30th Technical special committee on Fugen decommissioning which was held on September 19, 2014.
Imagawa, Yuya; Hashidate, Ryuta; Onizawa, Takashi; Kato, Shoichi; Otsuka, Satoshi; Onuma, Masato*; Nakashima, Hideharu*; Toyama, Takeshi*
no journal, ,
Oxide dispersion strengthened (ODS) steel, which has high strength at high to very high temperatures by uniform dispersion of nano-sized oxide particles with excellent thermal stability, is expected to be applied as a fuel cladding tube material to improve safety in the next generation sodium-cooled fast reactor (SFR). In this study, circumferential creep tests and evaluations of ODS steels at high to very high temperatures were carried out in order to obtain basic data for clarifying the correlation between microstructure and mechanical properties in the temperature range from high temperatures during abnormal transients to very high temperatures during accidents.
Ogawa, Fumio*; Imagawa, Yuya; Hashidate, Ryuta; Hiyoshi, Noritake*; Ito, Takamoto*
no journal, ,
In recent years, the demand for environmental protection has led to a trend toward shorter startup and shutdown intervals in thermal power plants. In fast breeder reactors (FBRs), more stringent operating environments (temperature and sodium environment) are also being considered. Modified 9Cr-1Mo steel is used in thermal power plants and fast breeder reactors because of its excellent mechanical properties at high temperatures. In actual machines, Modified 9Cr-1Mo steel is used in a creep-fatigue type with complex stress conditions, which may cause more damage than the uniaxial condition, so it is necessary to measure the multiaxial creep-fatigue life and establish a life evaluation method. In this study, axial and torsional multiaxial creep-fatigue tests were conducted in 550C air to accumulate property data and to evaluate the life. The results will be described later, but the strain rate and retention time dependence showed peculiarities different from those in the past. Next, the strain energy corresponding to the hysteresis loop area between stress and strain was calculated to evaluate the life, and the same energy was theorized to compare the actual and theoretical life. Specifically, a new life evaluation method based on strain energy was studied, which can evaluate the peculiar strain rate, retention time dependence, and multiaxial state observed in the proportional loading test.
Ogawa, Fumio*; Imagawa, Yuya; Hashidate, Ryuta; Wakai, Takashi; Hiyoshi, Noritake*; Ito, Takamoto*
no journal, ,
Multiaxial creep-fatigue tests in tension and torsion were performed on Mod.9Cr-1Mo steel at 550C. Specificity was observed with respect to strain rate and retention time dependence, indicating that creep-fatigue life is governed by the maximum stress and the degree of stress relaxation. The hysteresis energy was calculated and the lifetime was organized considering the effect of non-proportional loading. Specifically, the mechanism of hysteresis onset was examined, and the energy formulation was made considering the hardening index and relaxation stress. Although the life evaluation has been organized with coefficients of 2.9 and 2.2, this study was conducted with the goal of first organizing the life evaluation with a coefficient of 2.
Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Mitsuhara, Masatoshi*; Nakashima, Hideharu*; Onuma, Masato*; et al.
no journal, ,
no abstracts in English
Onizawa, Takashi; Toyota, Kodai; Imagawa, Yuya; Okajima, Satoshi; Ando, Masanori
no journal, ,
In order to realize a fast reactor that achieves both safety and economic efficiency at a high level, Japan Atomic Energy Agency (JAEA) is developing the material strength standard for fast reactor design. JAEA has developed the material strength standard based on the acquired data and its evaluation results, and the standard have been incorporated in the Japan Society of Mechanical Engineers (JSME) code, Rules on the Design and Construction of Nuclear Power Plants, Section II, Fast Reactors (JSME D&C FRs Code). This paper describes the standard that recently incorporated in the JSME D&C FRs code and ongoing studies for improvements in the near future.
Toyota, Kodai; Imagawa, Yuya; Onizawa, Takashi
no journal, ,
no abstracts in English
Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi
no journal, ,
no abstracts in English
