Kato, Tetsu*; Kawamura, Yamato*; Tahara, Junichiro*; Baba, Shoichiro*; Sanada, Yukihisa
Proceedings of the 30th (2020) International Ocean and Polar Engineering Conference (ISOPE 2020) (USB Flash Drive), p.1255 - 1260, 2020/10
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.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Hashidate, Ryuta; Kato, Shoichi; Kurihara, Akikazu
JAEA-Data/Code 2019-005, 117 Pages, 2019/07
SUS316 and SUS321 are used for structural materials of the Fast Breeder Reactors (FBRs), because of excellent high creep strength. This report summarized the mechanical properties data on SUS316 and SUS321 obtained in various tests including the long-term material tests and the material tests in sodium.
Yano, Yasuhide; Sekio, Yoshihiro; Tanno, Takashi; Kato, Shoichi; Inoue, Toshihiko; Oka, Hiroshi; Otsuka, Satoshi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; et al.
Journal of Nuclear Materials, 516, p.347 - 353, 2019/04
9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 C, which surpassed that of heat-resistant austenitic steels although creep rupture strength of tempered martensitic steels is generally lower than that of austenitic steels at high temperatures. The measured creep rupture strength of 9Cr-ODS steel claddings at 1000 C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 C. This superior strength seemed to be owing to transformation of the matrix from the -phase to the -phase. The transient burst strengths for 9Cr-ODS steel were much higher than those for 11Cr-ferritic/martensitic steel (PNC-FMS). Cumulative damage fraction analyses suggested that the life fraction rule can be used for the rupture life prediction of 9Cr-ODS steel and PNC-FMS claddings in the transient and accidental events with a certain accuracy.
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.
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.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation.
Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Inoue, Toshihiko; Kato, Shoichi; Furukawa, Tomohiro; Uwaba, Tomoyuki; Kaito, Takeji; Ukai, Shigeharu*; et al.
Journal of Nuclear Materials, 487, p.229 - 237, 2017/04
Ultra-high temperature ring tensile tests were carried out to investigate the tensile behavior of oxide dispersion strengthened (ODS) steel claddings and wrapper materials under severe accident conditions; temperatures ranged from room temperature to 1400C which is near the melting point of core materials. The experimental results showed that tensile strength of 9Cr-ODS steel claddings was highest in the core materials at the ultra-high temperatures between 900 and 1200C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200C. This degradation was attributed to grain boundary sliding deformation with / transformation, which was associated with reduced ductility. On the other hand, tensile strength of recrystallized 12Cr-ODS and FeCrAl-ODS steel claddings retained its high value above 1200 C unlike the other tested materials. Present study includes the result of "R&D of ODS ferritic steel fuel cladding for maintaining fuel integrity at the high temperature accident condition" entrusted to Hokkaido University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
Uwaba, Tomoyuki; Yano, Yasuhide; Otsuka, Satoshi; Naganuma, Masayuki; Tanno, Takashi; Oka, Hiroshi; Kato, Shoichi; Kaito, Takeji; Ukai, Shigeharu*; Kimura, Akihiko*; et al.
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04
Tolerance of fast rector fuel elements to failure in the typical accident conditions was evaluated for the oxide-dispersion-strengthened (ODS) ferritic steel claddings that are candidate of the cladding material for advanced fast reactors. The evaluation was based on the cladding creep damage, which was quantified by the cumulative damage fractions (CDFs). It was shown that the CDFs of the ODS ferritic steel cladding were substantially lower than the breach limit of 1.0 in the loss of flow and transient over power conditions until a passive reactor shutdown system operates.
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.
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 1.210/m, and for the capsule at a peripheral irradiation hole as 4.210/m. 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.
Shimomura, Kenta; Kato, Shoichi; Wakai, Takashi; Ando, Masanori; Hirose, Yuichi*; Sato, Kenichiro*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05
This paper describes experimental and analytical works to confirm that the design standard for SFR components sufficiently covers possible failure mechanisms. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. A series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. The curvature radii of the specimens were 1.6mm, 11.2mm and 40.0mm. The specimen having 1.6mm notch and 11.2mm notch failed from outer surface but the specimen having 40.0mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.
Furukawa, Tomohiro; Hirakawa, Yasushi; Kato, Shoichi; Iijima, Minoru; Otaka, Masahiko; Kondo, Hiroo; Kanemura, Takuji; Wakai, Eiichi
Fusion Engineering and Design, 89(12), p.2902 - 2909, 2014/12
For the irradiation test of the candidate materials for the fusion DEMO reactor, Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) are performed under the Broader Approach Activities. As a major Japanese activity on the target facility of the IFMIF, the engineering validation using the EVEDA Lithium Test Loop which is the largest scale liquid lithium test loop has been started in 2012. In parallel with the design and fabrication, the research on the technology establishment for the lithium safety handling was started in 2008, as one of the related technologies under the IFMIF-EVEDA. In the research, experiments of lithium chemical reaction, experiments on lithium fire, establishment of chemical analysis of impurities in lithium and experiments on advanced lithium leak detection system were carried out. This paper describes the results of these experiments.
Furukawa, Tomohiro; Hirakawa, Yasushi; Kato, Shoichi
Fusion Engineering and Design, 88(9-10), p.2502 - 2505, 2013/10
Lithium, which is used as the neutron source of the IFMIF, reacts with oxygen, nitrogen and moisture in atmosphere in case of the leakage accident. In this study, fundamental corrosion test was performed in order to obtain the corrosion behavior of austenitic stainless steel in the estimated lithium compounds. In the experiment, the lithium compounds were filled with the steel into the Tammann pipe made from AlO, and heated up to 850C. After the test, the specimen was cleaned by alcohol, and then the weight loss measurement and metallurgical examination were performed. Intense corrosion was observed at the environmental conditions containing lithium peroxide. No corrosion was observed in LiO environment. According to the consideration based on thermodynamics, LiO cannot oxidize iron and lithium is reducing agent. Slight corrosion was observed in LiOH and LiN environments.
Yoshida, Eiichi; Kato, Shoichi; Furukawa, Tomohiro
JAEA-Research 2012-034, 68 Pages, 2013/01
Oxygen concentration in sodium is the important factor for the corrosion of FBR structural materials. In this study, the experiments have been done to clarify the effect of sodium contaminated by oxygen on corrosion and low cycle fatigue strength of the materials. The materials for use of the experiments were FBR Grade type 316 SS (316FR) and Mod. 9Cr-1Mo steel. The corrosion test has been performed in sodium containing of 1, 10 and 10 ppm of initial oxygen at 650C for 500 hours. The fatigue test has been done for the post-immersed steels at 650C in air.
Furukawa, Tomohiro; Kato, Shoichi; Maeda, Shigetaka; Yamamoto, Masaya; Sekine, Takashi; Ito, Chikara
JAEA-Research 2011-039, 20 Pages, 2012/02
Application of zirconium alloy as a neutron reflector around the driver fuel region of the Japanese experimental fast reactor JOYO has been planned for a further increase of core average burn-up. In order to investigate the compatibility of the zirconium alloys with high-temperature sodium which is coolant of the JOYO, corrosion test in sodium and tensile test of the exposed alloys were performed. The corrosion test was done at 500C and 650C in stagnant/flowing sodium for two kinds of zirconium alloys, and then weight change measurement and metallurgical observation were carried out. The tensile test was performed in air at the same temperature with the sodium exposure.
Furukawa, Tomohiro; Kondo, Hiroo; Hirakawa, Yasushi; Kato, Shoichi; Matsushita, Izuru*; Ida, Mizuho; Nakamura, Kazuyuki
Fusion Engineering and Design, 86(9-11), p.2433 - 2436, 2011/10
In order to obtain the engineering data on the lithium target system, which is the neutron source of the International Fusion Material Irradiation Facility (IFMIF), the design and fabrication of the IFMIF/EVEDA Lithium Test Loop are being carried out under the Engineering Validation and Engineering Design Activity (EVEDA). The loop will hold 2.5 tons of lithium. Since lithium is specified by Japanese law as a dangerous substance, countermeasures which assumed a lithium leak incident and various abnormal issues are indispensable. This paper describes about the safety principles and measures for lithium leaks of the IFMIF/EVEDA lithium test loop decided under the detailed design process.
Furukawa, Tomohiro; Kato, Shoichi; Inagaki, Yoshiyuki; Aritomi, Masanori*
Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 5 Pages, 2011/10
A key problem in the application of a supercritical carbon dioxide (CO) turbine cycle to a fast breeder reactor is the corrosion of structural materials brought about by supercritical CO at high temperatures. In this study, high-temperature oxidation tests on the structural materials were performed in carbon dioxide pressurized at 0.2 and 1 MPa, and in air, and the oxidation behavior were compared. Results of investigating the effect of CO pressure including the previous reports tested at 10 MPa and at 20 MPa, the effect was hardly observed for all steels. In air environment, weight gain caused by high temperature oxidation was much lower than that in CO.