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JAEA Reports

FBR metallic materials test manual (2023 revised edition)

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.

Journal Articles

Validation of the applicability of the best-fit fatigue curves for 550$$^{circ}$$C in Mod.9Cr-1Mo steel to 1$$times$$10$$^{11}$$ cycles

Toyota, 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 1$$times$$10$$^{9}$$ cycles; to evaluate high cycle fatigue at 1$$times$$10$$^{9}$$ cycles, it is necessary to develop a best-fit fatigue curve applicable up to 1$$times$$10$$^{11}$$ 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 1$$times$$10$$^{11}$$ cycles was verified.

Journal Articles

Material data acquisition activities to develop the material strength standard for sodium-cooled fast reactors

Toyota, Kodai; Onizawa, Takashi; Wakai, Takashi; Hashidate, Ryuta; Kato, Shoichi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

JAEA Reports

Evaluation of tensile and creep properties on 9Cr-ODS steel claddings

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 850$$^{circ}$$C.

Journal Articles

Proposal of simulation material test technique for clarifying the structure failure mechanisms under excessive seismic loads

Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07

It is very essential to clarify the structure failure mechanisms under excessive seismic loads. However, structural tests using actual structural materials are very difficult and expensive. Therefore, we have proposed the structure test approach using lead alloys in order to simulate the structure failure mechanisms under the excessive seismic loads. In this study, we conducted material tests using lead alloy and verified the effectiveness of the simulated material tests. Moreover, we formulated inelastic constitutive equations (best fit fatigue curve equation and cyclic stress range - strain range relationship equation) of lead alloy based on the results of a series of material tests. Nonlinear numerical analyses, e.g. finite element analyses, can be performed using the proposed equations. A series of simulation material test technique enables structural tests and analyses using lead alloy to simulate the structure failure phenomena under excessive seismic loads.

Journal Articles

Proposal of simulation materials test technique and their constitutive equations for structural tests and analyses simulating severe accident conditions

Hashidate, Ryuta; Kato, Shoichi; Onizawa, Takashi; Wakai, Takashi; Kasahara, Naoto*

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08

Although it is very essential to clarify how the structure collapses under the severe accident conditions, the failure mechanisms in excessive high temperatures are not clarified. However, it is very difficult and expensive to perform structural tests using actual structural materials. Therefore, we propose to use lead alloys instead of actual structural materials. For demonstration of analogy between the failure mechanisms of lead alloys structure at low temperature and those of the actual structures at high temperature, numerical analyses are required. Although the authors proposed inelastic constitutive equations for numerical analyses in 2019, the equations could not successfully express because of large variations observed in the material tests of the lead alloy. In this study, we propose the improved inelastic constitutive equations of the lead alloy on the basis of the material test results used by aged alloy which can stabilized the material characteristic.

Journal Articles

Post irradiation experiment about SiC-coated oxidation-resistant graphite for high temperature gas-cooled reactor

Shibata, Taiju; Mizuta, Naoki; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; et al.

Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR). Oxidation damage on the graphite components in air ingress accident is a crucial issue for the safety point of view. SiC coating on graphite surface is a possible technique to enhance oxidation resistance. However, it is important to confirm the integrity of this material against high temperature and neutron irradiation for the application of the in-core components. JAEA and Japanese graphite companies carried out the R&D to develop the oxidation-resistant graphite. JAEA and INP investigated the irradiation effects on the oxidation-resistant graphite by using a framework of ISTC partner project. This paper describes the results of post irradiation experiment about the neutron irradiated SiC-coated oxidation-resistant graphite. A brand of oxidation-resistant graphite shows excellent performance against oxidation test after the irradiation.

Journal Articles

Enhancement of oxidation tolerance of graphite materials for high temperature gas-cooled reactor

Mizuta, Naoki; Sumita, Junya; Shibata, Taiju; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Sakaba, Nariaki

Tanso Zairyo Kagaku No Shinten; Nihon Gakutsu Shinkokai Dai-117-Iinkai 70-Shunen Kinen-Shi, p.161 - 166, 2018/10

To enhance oxidation resistance of graphite material for in-core components of HTGR, JAEA and four Japanese graphite companies; Toyo Tanso, IBIDEN, Tokai Carbon and Nippon Techno-Carbon, are carrying out for development of oxidation-resistant graphite by CVD-SiC coating. This paper describes the outline of neutron irradiation test about the oxidation-resistant graphite by WWR-K reactor of INP, Kazakhstan through an ISTC partner project. Prior to the irradiation test, the oxidation-resistant graphite by CVD-SiC coating of all specimens showed enough oxidation resistance under un-irradiation condition. The neutron irradiation test was already completed and out-of-pile oxidation test will be carried out at the hot-laboratory of WWR-K.

Journal Articles

Influence of cyclic softening on high temperature material properties in Mod.9Cr-1Mo steel

Onizawa, Takashi; Nagae, Yuji; Kato, Shoichi; Wakai, Takashi

Zairyo, 66(2), p.122 - 129, 2017/02

The applicability of Modified 9Cr-1Mo steel (ASME Grade 91 steel) as the main structural material in advanced loop-type sodium cooled fast reactor has been explored to enhance the safety, the credibility and the economic competitiveness of fast reactor plants. It is well-known that the steel exhibits cyclic softening behavior. Decrease of tensile and creep strength in softened materials has been already reported by other researchers. This paper discusses the relationship between cyclic softening conditions and high temperature material properties. Grade 91 steel was softened by repeat of plastic strain. The softening behavior could be evaluated by the index of the softening rate. Decrease of tensile and creep strength in softened materials can be evaluated by the softening rate and it depends on the cyclic softening conditions.

Journal Articles

Irradiation test about oxidation-resistant graphite in WWR-K research reactor

Shibata, Taiju; Sumita, Junya; Sakaba, Nariaki; Osaki, Takashi*; Kato, Hideki*; Izawa, Shoichi*; Muto, Takenori*; Gizatulin, S.*; Shaimerdenov, A.*; Dyussambayev, D.*; et al.

Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.567 - 571, 2016/11

Graphite are used for the in-core components of HTGR, and it is desirable to enhance oxidation resistance to keep much safety margin. SiC coating is the candidate method for this purpose. JAEA and four Japanese graphite companies are studying to develop oxidation-resistant graphite. Neutron irradiation test was carried out by WWR-K reactor of INP of Kazakhstan through ISTC partner project. The total irradiation cycles of WWR-K operation was 10 cycles by 200 days. Irradiation temperature about 1473 K would be attained. The maximum fast neutron fluence (E $$>$$0.18 MeV) for the capsule irradiated at a central irradiation hole was preliminary calculated as 1.2$$times$$10$$^{25}$$/m$$^{-2}$$, and for the capsule at a peripheral irradiation hole as 4.2$$times$$10$$^{24}$$/m$$^{-2}$$. Dimension and weight of the irradiated specimens were measured, and outer surface of the specimens were observed by optical microscope. For the irradiated oxidation resistant graphite, out-of-pile oxidation test will be carried out at an experimental laboratory.

JAEA Reports

Plan and reports of coupled irradiation (JRR-3 and JOYO of research reactors) and hot facilities work (WASTEF, JMTR-HL, MMF and FMF); R&D project on irradiation damage management technology for structural materials of long-life nuclear plant

Matsui, Yoshinori; Takahashi, Hiroyuki; Yamamoto, Masaya; Nakata, Masahito; Yoshitake, Tsunemitsu; Abe, Kazuyuki; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; et al.

JAEA-Technology 2009-072, 144 Pages, 2010/03

JAEA-Technology-2009-072.pdf:45.01MB

"R&D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant" was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of "Evaluation of Irradiation Damage Indicator" in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research-and-Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency.

Journal Articles

Long term efficiency and stability of MX precipitation strengthening of high chromium steel

Onizawa, Takashi; Ando, Masanori; Wakai, Takashi; Asayama, Tai; Kato, Shoichi

Tetsu To Hagane, 94(3), p.91 - 98, 2008/03

 Times Cited Count:3 Percentile:26.75(Metallurgy & Metallurgical Engineering)

Employment of high Cr steel as a main structural material is considered as a way to achieve economical competitiveness of fast breeder reactors. A series of trial products controlling V and Nb contents is produced and aging tests are conducted to investigate the long term stability of the MX strengthening mechanism. Before and after a long term aging process, metallurgical examinations and quantitative analyses are conducted to investigate the stability of MX particles. After aging, Z-phase was observed in high Cr steel content Nb. With precipitation and rapid coarsening of Z-phase, decrease in number of density of MX particles. Therefore, it is supposed that the long term efficiency of MX precipitation strengthening mechanisms is decrease in high Cr steel content Nb. In contrast, it is expected that VX precipitation strengthening mechanisms is stable in high Cr steel content only V, because Z-phase isn't precipitate and VX is stable after aging.

Journal Articles

Safeguards improvement for the Tokai Reprocessing Plant (TRP)

Onizawa, Toshikazu; Kimura, Takashi; Kurosu, Kazutoshi; Hayakawa, Tsuyoshi; Fukuhara, Junichi; Yatsu, Shoichi*

STI/PUB/1298 (CD-ROM), p.739 - 745, 2007/08

no abstracts in English

Journal Articles

$$^{52}$$Mn translocation in barley monitored using a positron-emitting tracer imaging system

Tsukamoto, Takashi*; Nakanishi, Hiromi*; Kiyomiya, Shoichiro*; Watanabe, Satoshi; Matsuhashi, Shimpei; Nishizawa, Naoko*; Mori, Satoshi*

Soil Science and Plant Nutrition, 52(6), p.717 - 725, 2006/12

 Times Cited Count:34 Percentile:63.07(Plant Sciences)

JAEA Reports

Material Properties of High Cr-Mo Steel (V) Mechanical properties of SUS410J3 Welded joint and base metal aged for 12000h

Onizawa, Takashi; Ando, Masanori; Kato, Shoichi; Yoshida, Eiichi

JNC TN9400 2005-019, 93 Pages, 2005/03

JNC-TN9400-2005-019.pdf:14.06MB

The high chromium (Cr) ferritic steels have both excellent thermal and mechanical properties at elevated temperature. Therefore, the applicability of high Cr ferritic steels for structural material of the future advanced Fast Breeder Reactor (FBR) is investigated. In this study, creep tests and metallurgical observations are conduced in order to investigate mechanical properties on SUS410J3 weld joint at FBR's temperature region. And tensile, impact tests and metallurgical observations after thermal aging for 12000 hours are conduced in order to evaluate mechanical properties and stability of microstructures on SUS410J3 material. As a result, following conclusions are obtained;Creep rupture properties up to about 3000 hours are obtained in weld metal zone at 823K and 873K. In this study, creep rupture strength show that the tendency of weld joint$$<$$ base metal (SR treatment) $$<$$ base metal$$<$$weld metal. Then the remarkable decreasing of creep fracture ductility was recognized at 873K for about 2000 hours. The possibilities of Type 4 damage are suggested because of the fracture position is fine HAZ region near the base metal from results of microstructure investigation etc. Therefore about the creep properties of a weld joint, long term creep dates are intend to expansion for evaluation. On the other hand dates of impact, tensile properties on high temperature and microstructure of stability about 350000 hours at 827K calculated by LMP are obtained. The decreasing of the impact properties by aging are saturated in 3000-6000 hours at 873K. And that decreasing rate is about 0.6-0.7 times at upper shelf energy (USE). Ultimate tensile strength and 0.2% proof stress are slightly decreases with time of aging. It is thought that reasons are inference of deposits, such as M$$_{23}$$C$$_{6}$$ etc.

JAEA Reports

Development of the high chromium ferritic steel FBR grade (2)Mechanical properties and observations of as received and aged some trial products controlling V and Nb contents

Onizawa, Takashi; Ando, Masanori; Wakai, Takashi; Kato, Shoichi; Aoto, Kazumi

JNC TN9400 2005-012, 99 Pages, 2005/03

JNC-TN9400-2005-012.pdf:9.77MB

The high chromium (Cr)ferritic steels have both excellent thermal properties and strength at elevated temperature. Therefore, the applicability of high Cr ferritic steels for structural material of the future advanced Fast Breeder Reactor is investigated. In previous study, a series of trial products controlling the vanadium (V)and niobium(Nb) contents is produced to investigate the efficiency and/or stability of these elements for the development of the high chromium ferritic steel FBR grade. In this study, tensile, impact, creep tests and observations are conducted in order to investigate the influence of V and Nb contents on the mechanical properties and stability of microstructures. These tests and observations are conducted for the as received materials and aged ones. As a result, the following conclusions are obtained ;(1)The higher strength and the lower ductility are obtained with the increases of V and Nb as far as this study investigates. There is little influence of V and Nb on relation between steady state creep rate, time to beginning of tertiary creep and creep rupture time.(2)The impact properties degrade with the increases of V and Nb, since Laves phase don't precipitate in the trial products with aging at 600 for 6000h. For that reason impact properties hardly fell.(3)Fine MX (V(C,N) and/or NbC) are observed in the steels which contain V and/or Nb. The number of MX increases with the increase of contents of V and/or Nb. It is show that they are likely to be stable under the condition of aging at 600 for 6000h.(4)The fine Cr$$_{2}$$ (C,N) is nucleated from V(C,N). And the fine Cr$$_{2}$$ (C,N) may be expected to suppress the growth of M$$_{23}$$C$$_{6}$$. This may be caused by V contained in Cr$$_{2}$$ (C,N).

JAEA Reports

Creep properties and microstructure change of FBR grade type 316 stainless steel weld zone

Hasebe, Shinichi; Onizawa, Takashi; Kato, Shoichi

JNC TN9400 2003-019, 62 Pages, 2003/03

JNC-TN9400-2003-019.pdf:3.33MB

We conducted a long-term creep test of weld metal and welded joint made from what we consider optimal filler metals for more than 10000 hours at 550$$^{circ}$$C, in order to evaluate their long-term properties at high temperatures, and select the appropriate filler metals for FBR grade type 316 stainless steel. We also conducted an evaluation of the long-term high temperature strength of the weld metals by observing changes in the microstructures that was subjected to material deterioration. The results obtained are as fo1lows. (1)Weld metal and welded joint made from 316 type and a specific optimal material for 16-8-2 type showed better creep properties than current materials. Especially for 16-8-2 type, the quality improved so much that predominance microstructure stability in the region of long-term. (2)We clearly showed that when $$delta$$-ferrite phase decomposed by long time heating. the Laves phase, $$sigma$$ phase and austenite phase were precipitated and the remaining $$delta$$ -ferrite phase was changed to an $$alpha$$-ferrite phase (Cr≒12%, Ni≒2%) as it became a low-alloy and reached equilibrium. (3)The long-term creep strength of the 316 type weld metal tends to decrease as $$sigma$$ phase separation increases due to a high Cr concentration in $$delta$$-ferrite phase. On the other hand, we confirmed that 16-8-2 type weld metal could maintain long-term creep strength almost as high as the base metal, because there was little separation of inter-metallic compounds due to its low concentration of Cr. (4)We found that change in the microstructure can be easily captured by analyzing the composition of the remaining $$delta$$-ferrite phase. This is an effective method to evaluate long-term high temperature strength.

JAEA Reports

Material properties of high Cr-Mo steel (II); Physical properties of HCM12A(2001)

Kato, Shoichi; Onizawa, Takashi; Yoshida, Eiichi

JNC TN9400 2003-015, 31 Pages, 2003/03

JNC-TN9400-2003-015.pdf:4.62MB

A high Cr-Mo steel is candidate for a structural materials of future LMFBR, because of good thermal properties and high creep strength. In this study, material physical tests were carried out on the HCM12A (2001) of high Cr-Mo steel in order to evaluation of the basic physical properties. This report only shows the measurement results of the physical properties data of the rolled steel plates (2 heats) but doesn't give the design standards of the physical properties for the future LMFBR. The evaluated physical properties are as follows ; (1) Specific gravity (R.T), (2)Specific heat (R.T.$$sim$$900$$^{circ}$$C), (3) Thermal conductivity coefficient (R.T.$$sim$$900$$^{circ}$$C), (4) Thermal expansion coefficient (50$$^{circ}$$C$$sim$$1000$$^{circ}$$C), (5) Young's modulus (R.T.$$sim$$700$$^{circ}$$C), (6) Poisson's ratio (R.T.$$sim$$700$$^{circ}$$C)

JAEA Reports

FBR Metallic materials test manual (revised edition)

Kato, Shoichi; Onizawa, Takashi; Yoshida, Eiichi; Aoto, Kazumi

JNC TN9520 2001-001, 116 Pages, 2001/01

JNC-TN9520-2001-001.pdf:3.44MB

For the development of the fast breeder reactor, this manual described the method of in-air and in-sodium test for materials and the arrangement method of the data. This manual is reflected the revision of Japanese Industrial Standard (JIS) and change to the international unit. The test method of domestic committees such as the VAMAS (Versailles Project on Advanced Materials and Standards) workshop was also considered. And, the material test technologies accumulated in this group until now were arranged in order to transmit all of them.

JAEA Reports

Outline of a fuel treatment facility in NUCEF

Sugikawa, Susumu; ; ; Nakazaki, Masato; Shirahashi, Koichi; ; *; *; Tsuji, Kenichi*; Tachimori, Shoichi; et al.

JAERI-Tech 97-007, 86 Pages, 1997/03

JAERI-Tech-97-007.pdf:3.27MB

no abstracts in English

33 (Records 1-20 displayed on this page)