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C to 1000CImagawa, Yuya; Hashidate, Ryuta; Miyazawa, Takeshi; Onizawa, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 61(6), p.762 - 777, 2024/06
The Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650C to 850C. However, little data have been obtained above 850C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700C to 1000C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix's phase transformation, and a single equation can express a creep rupture strength from 700C to 1000C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.
Department of Materials Science; Department of Fusion Engineering Research (Tokai Site)
JAERI-Review 2004-018, 97 Pages, 2004/08
JAERI-Review-2004-018.pdf:18.92MB
Extensive efforts for evaluating the irradiation performances of a reduced activation ferritic/martensitic steel (RAF/M) of F82H* and other several RAF/Ms have been made in recent several years. They are, examinations of the effects of neutron irradiation on (1) Ductile to brittle transition temperature (DBTT) up to a damage level of 20 dpa to explore lower temperature limit, (2) Enhanced He effect on DBTT shift for Ni/B doped heats (isotopic tailoring method was used for B doping), (3) Susceptibility to environmentally assisted cracking by the slow strain rate tensile tests (SSRT) in a high temperature pressurized water and (4) Flow stress-plastic strain relation obtained by measuring the profile of the specimen during tensile testing, together with the activities of (5) the development of the test methods after neutron irradiation and (6) other supporting researches. Results are summarized in the present report. They clearly indicate the good applicability of RAF/Ms to fusion machines.
; Yoshida, Eiichi; Aoto, Kazumi
JNC TN9400 2000-042, 112 Pages, 2000/03
JNC-TN9400-2000-042.pdf:8.55MB
A sodium-water reaction drove from the single tube break in steam generator of FBR might overheat labor tubes rapidly under internal pressure loadings. lf the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. This study clarified the tensile and creep properties of Mod.9Cr-1Mo steel at ultra-high temperature which will be used in evaluation of the tube burst by sodium-water reaction. The strain rates for tensile test are from 10%/min to 10%/sec, and creep-rupture time is maximum 277sec. The range of test temperature is 700C to 1300C. The main results obtained were as follows; (1)The evaluation data on the relationship between tensile strength and strain rate and creep-rupture strength in shorter time on Mod.9Cr-1Mo steel were acquired. (2)Short-term mechanical properties of Mod.9Cr-1Mo steel were evaluated based on the results of tensile and creep-rupture tests up to 1300C. As a result of the evaluation, recommended equation of creep-rupture strength in the short-term was proposed. (3)Tensile and creep-rupture strength of Mod.9Cr-1Mo steel tube showed the value which was higher than the 2 1/4Cr-1Mo steel, and it was proven to have the superior properties.
; ; ; ; Sago, Hiromi*; ;
JNC TJ8400 2000-049, 161 Pages, 2000/02
JNC-TJ8400-2000-049.pdf:9.56MB
In this study basic data on welds of overpack structures for HLW were acquired and a predictive destruction analysis was performed usig the data acquired, in order to examine the viability of weld design methods. The results are summarized as follows: (1)Investigation of Design and Welding Condition for Welded Joint Models. Three welding methods--EBW, TIG and MAG--were selected, and welding conditions were determined so that the welding quality almost equivalent to that of an actual over-pack was ensured. (2)Fabrication of Welded Joint Models. Three welded joint models, one for each of EBW, TIG and MAG, were fabricated. It was confirmed that these models satisfied the quality requirements for Class I specified in JIS Z3104. (3)Sampling and Machining of Strength Test Specimens. Test specimens were taken from each welded joint model, and models for corrosion tests were delivered to the Japan Nuclear Cycle Development Institute (JNC). (4)Strength Test and Micro/macro Structure observation. Tensile tests were conducted at room temperature and at 150C, and fracture toughness tests at 0C and 150C, in order to obtain stress-strain curves, J-R curves and Vickers hardness. In addition, an observation of micro and macro structures was performed. (5)Evaluation. Using the data on the welds obtained from the tests, a fracture prediction analysis and an evaluation of unstable fracture due to weld flaws were performed on the over-pack design described in the second progress report. The following conclusions were obtained: (a)For the overpack design examined, the effects of welds (material property and residual stress) and fabrication tolerance on fracture loading are negligible. (b)In addition, it was decided that even in a design with reduced wall thickness, welds have an insignificant effect on fracture loading because fracture initiates in the center of the shell of the overpack. (c)The size of flaws leading to unstable fracture is on ...
; ; ; ; Sago, Hiromi*; ;
JNC TJ8400 2000-048, 30 Pages, 2000/02
JNC-TJ8400-2000-048.pdf:1.64MB
In this study basic data on welds of overpack structures for HLW were acquired and a predictive destruction analysis was performed using the data acquired, in order to examine the viability of weld design methods. The results are summarized as follows: (1)Investigation of Design and Welding Conditions for Welded Joint Models. Three welding methods--EBW, TIG and MAG-were selected, and welding conditions were determined so that the welding quality almost equivalent to that of an actual over-pack was ensured. (2)Fabrication of Welded Joint Models. Three welded joint models, one for each of EBW, TIG and MAG, were fabricated. It was confirmed that these models satisfied the quality requirements for Class I specified in JIS Z3104. (3)Sampling and Machining of Strength Test Specimens. Test specimens were taken from each welded joint model, and models for corrosion tests were delivered to the Japan Nuclear Cycle Development Institute (JNC). (4)Strength Test and Micro/macro Structure observation. Tensile tests were conducted at room temperature and at 150C, and fracture toughness tests at 0C and 150C, in order to obtain stress-strain curves, J-R curves and Vickers hardness. In addition, an observation of micro and macro structures was performed. (5)Evaluation. Using the data on the welds obtained from the tests, a fracture prediction analysis and an evaluation of unstable fracture due to weld flaws were performed on the over-pack design described in the second progress report. The following conclusions were obtained: (a)For the overpack design examined, the effects of welds (material property and residual stress) and fabrication tolerance on fracture loading are negligible. (b)In addition, it was decided that even in a design with reduced wall thickness, welds have an insignificant effect on fracture loading because fracture initiates in the center of the shell of the overpack. (c)The size of flaws leading to unstable fracture is on the ...
Watanabe, Katsutoshi; Nakajima, Hajime; Sahira, Kensho*; Marushichi, Koki*; Takeiri, Toshiki*; Saito, Teiichiro*; Takatsu, Tamao*; Nakanishi, Tsuneo*; Koikegami, Hajime*; Higuchi, Makoto*
Proc. of the 2nd JAERI Symp. on HTGR Technologies; Oct. 21
23,1992,Oarai,Japan (JAERI-M 92-215), p.560 - 574, 1993/01
no abstracts in English
Watanabe, Katsutoshi; Sahira, Kensho*; Takatsu, Tamao*; Nakanishi, Tsuneo*; Higuchi, Makoto*; Nakajima, Hajime
Computer Aided Innovation of New Materials,II, p.1491 - 1494, 1993/00
no abstracts in English
Watanabe, Katsutoshi; Nakajima, Hajime; Sahira, Kensho*; Marushichi, Koki*; Takeiri, Toshiki*; Saito, Teiichiro*; Takatsu, Tamao*; Nakanishi, Tsuneo*; Koikegami, Hajime*; Higuchi, Makoto*
Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 33(2), p.227 - 242, 1992/00
no abstracts in English
Watanabe, Katsutoshi; Nakajima, Hajime; Sahira, Kensho*; Marushichi, Koki*; Takeiri, Toshiki*; Saito, Teiichiro*; Takatsu, Tamao*; Nakanishi, Tsuneo*; Koikegami, Hajime*; Higuchi, Makoto*
JAERI-M 91-189, 67 Pages, 1991/11
no abstracts in English
; ; ; ; ; Yoshida, Eiichi;
PNC TN9450 91-010, 259 Pages, 1991/10
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the creep properties of Mod.9Cr-1Mo steels for steam generator, based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; Material : Mod.9Cr-1Mo steels (Base Metal) Plate 7 Heats (F2, F6, F7, F9, F10, NSC1, NSC2) Forging 8 Heats (F4, F5, F8, F11, VIM, ESR, F520, F550) Tube 1 Heats (F3) Test temperature : 450650C Test method : According to JIS and FBR Metallic Materials Test Method Test environment : In Air and in Sodium Number of deta : 314 points
; Hirakawa, Yasushi; Furukawa, Tomohiro; ; ; ;
PNC TN9450 91-004, 71 Pages, 1991/07
In order to advancement in materials strength standard on elevated temperature design guide of the FBRs and evaluation method of materials strength behavior, this report are presented about the low-cycle fatigue properties of Mod.9Cr-1Mo steel, based on the R&D results obtained through the sctivities of material tests. Contents of the data sheet are as follows; [Material ; Mod.9Cr-1Mo steel(SR)] F2 Heat 1,000
1,00012mm
(Plate) F4 Heat 1,0001,000250mm(Forging) F6 Heat 1,0001,00025mm(Plate) [Environment; In Air and in Sodium] [Test temperature ; 450, 500, 550, 600 and 650C] [Strain rate ; 0.1%/sec (10
mm/mm/sec)] [Strain range ; 0.38% 1.86%] [Number of deta ; 83 points]
Watanabe, Katsutoshi; Nakajima, Hajime; Sahira, Kensho*; ; Takeiri, Toshiki*; Saito, Teiichiro*; Takatsu, Tamao*; Nakanishi, Tsuneo*
Journal of Nuclear Materials, 185, p.8 - 18, 1991/00
Times Cited Count：2 Percentile：31.86(Materials Science, Multidisciplinary)no abstracts in English
Watanabe, Katsutoshi; Nakanishi, Tsuneo*; Takatsu, Tamao*; Sahira, Kensho*; Nakajima, Hajime
Koon Gakkai-Shi, 16(6), p.368 - 376, 1990/11
no abstracts in English
Kondo, Tatsuo; ; ; ; ; ; ; ;
JAERI-M 86-003, 228 Pages, 1986/02
no abstracts in English
Wakai, Eiichi; Shibayama, Tamaki*; Noto, Hiroyuki*; Wakui, Takashi
no journal, ,
In this study, we fabricated a prototype iron-based high-entropy alloy (Fe-Mn-V-Cr-Al-C) composed of low activation elements (free of Ni and Co) by radio frequency melting method and evaluated its basic properties, aiming to apply it to new functional materials for high-energy accelerator target system components, nuclear reactors and fusion reactors. XRD analysis of this material revealed that it has a BCC crystal structure, in which vanadium carbide (VC) is precipitated. This material was found not only to be magnetic, but also to have a fairly small magnetic domain structure. As for the grain size, relatively small grains (about 20-50 
m) were observed despite the homogenization heat treatment at 1150C. With regard to strength properties, the results of Vickers hardness measurements indicated that the alloy was much harder than ordinary iron alloys, slightly exceeding pure tungsten, and elastic wave velocity measurements showed that it was faster than iron-based materials and had a higher elastic modulus than stainless steel.
Wakai, Eiichi; Noto, Hiroyuki*; Shibayama, Tamaki*; Nakagawa, Yuki*; Ishida, Taku*; Makimura, Shunsuke*; Wakui, Takashi; Furuya, Kazuyuki*; Ando, Masami*
no journal, ,
In the fields of energy, nuclear power, high-energy accelerator target systems, nuclear fusion, and biology, radiation causes degradation of materials and equipment, and thus it is expected to create new materials with high durability and superior functionality. In this study, for Fe-, Ti-, and W-based high-entropy alloys (HEA) composed of low activation elements (Ni and Co free), Fe-based alloys were prepared by radio frequency melting, Ti-based alloys by cold crucible levitation melting, and W-based alloys by arc melting using metal powders. These materials were tested by X-ray diffraction, microstructural observation, hardness measurement, magnetic measurement, electrical resistivity measurement, scanning transmission electron microscope STEM (or TEM, SEM) and energy dispersive X-ray spectroscopy, ultrasonic measurement, and hot isostatic pressing (HIP) method. These HEAs were found to be much harder than normal alloys, and in Fe-based HEAs, the magnetic properties and related microstructural analysis showed that they have interesting characteristics such as micro magnetic domain structures. In particular, for Fe- and W-based HEAs, the changes in crystal structure, orientation, and internal microstructure caused by HIP treatment and the accompanying effects of high temperature and pressure have been found to have a significant effect on magnetic properties and material strength properties.
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
Tanigawa, Hiroyasu*; Nakajima, Motoki*; Sakasegawa, Hideo; Ando, Masami*; Watanabe, Yoshiyuki*; Nozawa, Takashi*; Kato, Yutai*; Nakata, Toshiya*
no journal, ,
Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Tokita, Shun*; Fujii, Hiromichi*; Sato, Yutaka*; Kokawa, Hiroyuki*
no journal, ,
no abstracts in English
Shobu, Takahisa
no journal, ,
no abstracts in English
