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Journal Articles

High-temperature creep properties of 9Cr-ODS tempered martensitic steel and quantitative correlation with its nanometer-scale structure

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:2 Percentile:82.38(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 $$^{circ}$$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 $$^{circ}$$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.

Journal Articles

High temperature mechanical properties and microstructure in 9Cr or 12Cr oxide dispersion strengthened steels

Mitsuhara, Masatoshi*; Kurino, Koichi*; Yano, Yasuhide; Otsuka, Satoshi; Toyama, Takeshi*; Onuma, Masato*; Nakashima, Hideharu*

Tetsu To Hagane, 109(3), p.189 - 200, 2023/03

 Times Cited Count:0 Percentile:0(Metallurgy & Metallurgical Engineering)

Oxide Dispersion Strengthened (ODS) ferritic steel, a candidate material for fast reactor fuel cladding, has low thermal expansion, good thermal conductivity, and excellent resistance to irradiation damage and high temperature strength. The origin of the excellent high-temperature strength lies in the dispersion of fine oxides. In this study, creep tests at 700 or 750$$^{circ}$$C, which are close to the operating temperatures of fast reactors, and high-temperature tensile tests at 900 to 1350 $$^{circ}$$C, which simulate accident conditions, were conducted on 9Cr ODS ferritic steels, M11 and MP23, and 12Cr ODS ferritic steel, F14, to confirm the growth behavior of oxides. In the M11 and F14 creep test samples, there was little oxide growth or decrease in number density from the initial state, indicating that dispersion strengthening by oxides was effective during deformation. After creep deformation of F14, the development of dislocation substructures such as dislocation walls and subgrain boundaries was hardly observed, and mobile dislocations were homogeneously distributed in the grains. The dislocation density increased with increasing stress during the creep test. In the high-temperature ring tensile tests of MP23 and F14, the strength of both steels decreased at higher temperatures. In MP23, elongation decreased with increasing test temperature from 900 to 1100 $$^{circ}$$C, but increased at 1200 $$^{circ}$$C, decreased drastically at 1250 $$^{circ}$$C, and increased again at 1300 $$^{circ}$$C. In F14, elongation decreased with increasing temperature. It was inferred that the formation of the $$delta$$-ferrite phase was responsible for this complex change in mechanical properties of MP23 from 1200 to 1300 $$^{circ}$$C.

Journal Articles

Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Science and Technology, 9 Pages, 2023/00

 Times Cited Count:0 Percentile:79.09

The effects of fast neutron irradiation on tensile properties of modified 316 stainless steel (PNC316) claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400 and 735 $$^{circ}$$C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400 and 500$$^{circ}$$C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Journal Articles

Effect of nitrogen concentration on creep strength and microstructure of 9Cr-ODS ferritic/martensitic steel

Oka, Hiroshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Hashimoto, Naoyuki*

Journal of Nuclear Materials, 572, p.154032_1 - 154032_8, 2022/12

 Times Cited Count:2 Percentile:73.56(Materials Science, Multidisciplinary)

9Cr oxide dispersion strengthened steels with slightly different nitrogen concentrations (0.0034 - 0.029 wt%) were prepared and their creep property at 973 K was investigated with microstructural characterization before and after the creep test. The creep strength decreased significantly as the nitrogen concentration increased. Microstructural observation revealed that, in the higher nitrogen concentration specimen, coarse Y-rich inclusions were found along the boundary between transformed ferrite region and residual ferrite region. The solubility difference of nitrogen in $$alpha$$ and $$gamma$$ phase would induce the localized increment of nitrogen concentration in the boundary region during the austenitizing process, resulting in the thermodynamic destabilization and subsequent coarsening of the dispersed oxide particles. The rows of creep voids were found near the rupture part of the crept specimen, suggesting that the coarse inclusions were the starting point of creep void formation and the subsequent premature fracture.

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

Tensile properties on dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging

Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Materials, 555, p.153105_1 - 153105_8, 2021/11

 Times Cited Count:0 Percentile:17.78(Materials Science, Multidisciplinary)

The aim of this study was to evaluate the tensile properties and microstructures of dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging at temperatures between 400 and 600$$^{circ}$$C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy and transmission electron microscopy. Microstructural analysis showed that the microstructure in the weld metals consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450$$^{circ}$$C due to the effects of both a-a' phase separation and G-phase precipitation. However, there was no significant change in the total elongation, and fracture surfaces indicated that very fine dimpled rupture was predominant rather than the cleavage rupture. It was suggested that lath martensite phases enhanced the tensile strength due to phase separation, while residual austenite played a role in keeping elongation as a soft phase.

Journal Articles

Laser beam direct energy deposition of graded austenitic-to-martensitic steel junctions compared to dissimilar electron beam welding

Villaret, F.*; Boulnat, X.*; Aubry, P.*; Yano, Yasuhide; Otsuka, Satoshi; Fabregue, D.*; de Carlan, Y.*

Materials Science & Engineering A, 824, p.141794_1 - 141794_10, 2021/09

 Times Cited Count:2 Percentile:28.17(Nanoscience & Nanotechnology)

Journal Articles

Effects of thermal aging on the mechanical properties of FeCrAl-ODS alloy claddings

Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*

Materials Transactions, 62(8), p.1239 - 1246, 2021/08

 Times Cited Count:3 Percentile:36.78(Materials Science, Multidisciplinary)

The FeCrAl-ODS alloy claddings were manufactured and Vickers hardness, ring tensile tests and TEM observations of these claddings were performed to investigate the effects of thermal aging at 450 $$^{circ}$$C for 5,000 and 15,000 h. The age-hardening of all FeCrAl-ODS alloy cladding was found. In addition, the significant increase in tensile strength was accompanied by much larger loss of ductility. It was suggested that this age-hardening behavior was attributed to the (Ti, Al)-enriched phase ($$beta$$' phase) and the $$alpha$$' phase precipitates (content of Al is $$<$$ 7 wt%). In comparison with FeCrAl-ODS alloys with almost same chemical compositions, there was significant age-hardening in both alloys. However, the extrusion bar with no-recrystallized structures was keeping good ductility. It was suggested that this different behavior of reduction ductility was attributed to the effects of grain boundaries, dislocation densities and specimen preparation direction.

Journal Articles

Microstructural stability of ODS steel after very long-term creep test

Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki

Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04

 Times Cited Count:5 Percentile:74.69(Materials Science, Multidisciplinary)

In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700$$^{circ}$$C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.

Journal Articles

Solid-solution strengthening by Al and Cr in FeCrAl oxide-dispersion-strengthened alloys

Ukai, Shigeharu*; Yano, Yasuhide; Inoue, Toshihiko; Sowa, Takashi*

Materials Science & Engineering A, 812, p.141076_1 - 141076_11, 2021/04

 Times Cited Count:11 Percentile:76.11(Nanoscience & Nanotechnology)

FeCrAl oxide dispersion strengthened alloys are promising materials for accident tolerant fuels for light water reactors (LWRs). In these alloys, Al and Cr are key elements with important synergistic effects: enhancement of the formation of oxidation-resistant Al$$_{2}$$O$$_{3}$$ phase by Cr addition and suppression of the formation of the embrittling Cr-rich $$alpha$$' phase by Al addition. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The solid-solution strengthening resulting from Al and Cr co-addition was investigated in this study. The Al and Cr contents were systematically varied from 9-16 at.% and 10-17 at.%, respectively, and tensile tests were conducted at 298 K, 573 K and 973 K in the as-annealed condition. The solid solution strengthening increased linearly, 20 MPa per 1 at.% Al and 5 MPa per 1 at.% Cr, at the typical LWR operational temperature of 573 K. The conventional Fleischer-Friedel and Labusch theories cannot explain this level of solid-solution strengthening. It was shown that Suzuki's double kink theory for screw dislocations reasonably predicts the solid solution strengthening by Al and Cr as well as the inverse dependency on the absolute temperature and linear dependency on the Al and Cr content.

Journal Articles

Development of ODS tempered martensitic steel for high burn up fuel cladding tube of SFR

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.

Journal Articles

Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

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

 Times Cited Count:13 Percentile:85.48(Materials Science, Multidisciplinary)

9Cr-ODS steel claddings consisting of tempered martensitic matrix, showed prominent creep rupture strength at 1000 $$^{circ}$$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 $$^{circ}$$C was higher than that from extrapolated creep rupture trend curves formulated using data at temperatures from 650 to 850 $$^{circ}$$C. This superior strength seemed to be owing to transformation of the matrix from the $$alpha$$-phase to the $$gamma$$-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.

Journal Articles

Effect of nitrogen concentration on nano-structure and high-temperature strength of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji

Nuclear Materials and Energy (Internet), 16, p.230 - 237, 2018/08

 Times Cited Count:4 Percentile:39.44(Nuclear Science & Technology)

Journal Articles

Model calculation of Cr dissolution behavior of ODS ferritic steel in high-temperature flowing sodium environment

Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji

Journal of Nuclear Materials, 505, p.44 - 53, 2018/07

AA2017-0603.pdf:1.7MB

 Times Cited Count:2 Percentile:21.58(Materials Science, Multidisciplinary)

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.

Journal Articles

Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions

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

 Times Cited Count:32 Percentile:96.99(Materials Science, Multidisciplinary)

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 1400$$^{circ}$$C 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 1200$$^{circ}$$C, but that there was significant degradation in tensile strength of 9Cr-ODS steel claddings above 1200$$^{circ}$$C. This degradation was attributed to grain boundary sliding deformation with $$gamma$$/$$delta$$ 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 $$^{circ}$$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).

Journal Articles

Evaluation on tolerance to failure of ODS ferritic steel claddings at the accident conditions of fast reactors

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.

Journal Articles

Higher harmonic imaging of small defects in ODS steel cladding tubes and characterization of the defects with SEM

Kawashima, Koichiro*; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji

Dai-24-Kai Choompa Ni Yoru Hihakai Hyoka Shimpojiumu Koen Rombunshu (USB Flash Drive), p.99 - 104, 2017/01

no abstracts in English

Journal Articles

Effect of thermo-mechanical treatments on nano-structure of 9Cr-ODS steel

Oka, Hiroshi; Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Uwaba, Tomoyuki; Kaito, Takeji; Onuma, Masato*

Nuclear Materials and Energy (Internet), 9, p.346 - 352, 2016/12

 Times Cited Count:21 Percentile:89.26(Nuclear Science & Technology)

Journal Articles

Tensile properties and hardness of two types of 11Cr-ferritic/martensitic steel after aging up to 45,000 h

Yano, Yasuhide; Tanno, Takashi; Sekio, Yoshihiro; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.324 - 330, 2016/12

BB2015-1728.pdf:1.04MB

 Times Cited Count:14 Percentile:79.83(Nuclear Science & Technology)

Journal Articles

Strength anisotropy of rolled 11Cr-ODS steel

Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.353 - 359, 2016/12

BB2015-1727.pdf:6.74MB

 Times Cited Count:9 Percentile:65.49(Nuclear Science & Technology)

Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must be excellent in high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Oxide dispersion strengthened (ODS) steels have been developing as the candidate material. Japan Atomic Energy Agency (JAEA) have been developing 9 and 11 Chromium (Cr) ODS steels for advanced fast reactor cladding tubes. The JAEA 11Cr-ODS steels were rolled in order to evaluate their anisotropy. Tensile tests and creep tests of them were carried out at 700 $$^{circ}$$C in longitudinal and transverse orientation. The anisotropy of tensile strength was negligible, though that of creep strength was distinct. The observation results and chemical composition analysis suggested that the cause of the anisotropy in creep strength was prior powder boundary including Ti-rich precipitates.

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