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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:1 Percentile:58.8(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:0 Percentile:100(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:1 Percentile:75.54(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:17 Percentile:3.04(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:10 Percentile:18.31(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:7 Percentile:31.55(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:4 Percentile:47.23(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.

Journal Articles

Mechanical properties and microstructure of dissimilar friction stir welds of 11Cr-ferritic/martensitic steel to 316 stainless steel

Sato, Yutaka*; Kokawa, Hiroyuki*; Fujii, Hiromichi*; Yano, Yasuhide; Sekio, Yoshihiro

Metallurgical and Materials Transactions A, 46(12), p.5789 - 5800, 2015/12

Dissimilar friction stir welding (FSW) of an 11% Cr ferritic/martensitic stee (PNC-FMS) to 316-grade austenitic stainless steel was attempted with a view to its potential application to the wrapper tubes of next-generation fast reactors. The mechanical properties and microstructure of the resulting welds were systematically examined, which revealed that FSW produces a defect-free stir zone in which material intermixing is notably absent. That is, both steels are separately distributed along a zigzagging interface in the stir zone when PNC-FMS is placed on the retreating side, with the tool plunging at the butt line. This interface did not act as a fracture site during small-sized tensile testing of the stir zone. Furthermore, the microstructure of the stir zone was refined in both the PNC-FMS and 316 stainless steel sides, resulting in improved mechanical properties over the respective base material regions.

Journal Articles

Weldability of dissimilar joint between PNC-FMS and Type 316 steel under electron beam welding

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

Journal of Nuclear Science and Technology, 52(4), p.568 - 579, 2015/04

 Times Cited Count:2 Percentile:74.39(Nuclear Science & Technology)

The dissimilar butt welding joint of 11Cr-ferritic/martensitic steel (PNC-FMS) and Type 316 austenitic steel (SUS316) produced by electron beam (EB) welding was studied. This study was carried out to investigate optimization of EB welding and post weld heat treatment (PWHT). Optimum EB welding conditions were a focus position of 30-40 mm and a welding speed of 1750-2000 mm/min, and optimum PWHT was performed after welding at 690$$^{circ}$$C for 60 min. As a result, no formation of delta ferrite was observed adjacent to the fusion zone, and the mechanical properties of the welds were similar to those of the base material. In this regard, EB welding is a proper fusion welding process for dissimilar PNC-FMS and SUS316.

Journal Articles

Effects of manufacturing process on impact properties and microstructures of ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Tanaka, Kenya

Journal of Nuclear Materials, 455(1-3), p.480 - 485, 2014/12

 Times Cited Count:10 Percentile:30.51(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) steels are noticed as an advanced alloy durable to high-temperature and high-dose neutron irradiation environment. Japan Atomic Energy Agency, 9-12Cr-ODS martensite steels have been developed as the primary candidate material for fast reactor fuel cladding tube. They would be also good candidates for fusion reactor blanket material. In this work, two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Tensile tests, creep tests, 1/3 sized Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steel was shown to be much superior than that of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores). The ductility of full pre-alloy ODS steels were better than that of pre-mix ODS steels.

JAEA Reports

Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.

JAEA-Research 2013-030, 57 Pages, 2013/11

JAEA-Research-2013-030.pdf:48.2MB

It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835$$^{circ}$$C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Journal Articles

Microstructure characterization of oxide dispersion strengthened steels containing metallic chromium inclusions after high-temperature thermal aging

Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Tanaka, Kenya

Materials Transactions, 54(10), p.2018 - 2026, 2013/10

 Times Cited Count:4 Percentile:65.99(Materials Science, Multidisciplinary)

Microstructure characterizations of 9Cr-oxide dispersion strengthened (ODS) steels were carried out after high-temperature thermal aging to reproduce the anomalous microstructure change that occurred in the BOR-60 irradiation test-formation of abnormally coarse and irregular precipitates a few tens of micrometers in size. In the 9Cr-ODS steel containing metallic Cr inclusions, coarse and irregular precipitates were formed nearby metallic Cr inclusions after the 750$$^{circ}$$C thermal aging for 8,000h. Based on the analyses using energy dispersive X-ray spectrometry (EDX) and electron backscattered pattern (EBSP), coarse and irregular precipitates were identified as M23C6.

Journal Articles

Mechanical properties of friction stir welded 11Cr-ferritic/martensitic steel

Yano, Yasuhide; Sato, Yutaka*; Sekio, Yoshihiro; Otsuka, Satoshi; Kaito, Takeji; Ogawa, Ryuichiro; Kokawa, Hiroyuki*

Journal of Nuclear Materials, 442(1-3), p.S524 - S528, 2013/09

 Times Cited Count:10 Percentile:28.98(Materials Science, Multidisciplinary)

Friction stir welding was applied to the wrapper tube materials, 11Cr-ferritic/martensitic steel, intended for fast reactors and defect-free welds were successfully produced. Then, the mechanical and microstructural properties of the friction stir welded steel were investigated. The hardness values of the stir zone were about 550 Hv, and they had hardly any dependence on the rotational speed, although they were much higher than that of the base material. However, tensile strengths and elongations of the stir zones were better at 298 K, compared to those of the base material. These excellent tensile properties were attributable to the fine grain formation during friction stir welding. A part of this study is the result of "Friction stir welding of the wrapper tube materials for Na fast reactors" carried out under the Strategic Promotion Program for Basic Nuclear Research by the Ministry of Education, Culture, Sports, Science and Technology of Japan.

Journal Articles

Microstructure and high-temperature strength of high Cr ODS tempered martensitic steels

Otsuka, Satoshi; Kaito, Takeji; Tanno, Takashi; Yano, Yasuhide; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 442(1-3), p.S89 - S94, 2013/09

 Times Cited Count:11 Percentile:25.84(Materials Science, Multidisciplinary)

The manufacturing tests of 11-12Cr ODS tempered martensitic steels were carried out, and their ferritic/martensitic duplex structures were quantitatively evaluated by three types of methods, i.e. high temperature XRD, EPMA and metallography. It was demonstrated that excessive formation of residual-alpha ferrite provided by increasing Cr can be suppressed by appropriately controlling the concentration of ferrite-forming element and austenite-forming element on the basis of the parameter "chemical driving force of $$alpha$$ to $$gamma$$ reverse transformation" as a useful indication. The 11Cr-ODS steel containing a small portion of residual-alpha ferrite was successfully manufactured. In the as-received condition, this 11Cr-ODS steel is shown to have the satisfactory creep strength and ductility as high as the 9Cr-ODS steel while 0.2% proof strength at 973K is lower than in the 9Cr-ODS steel.

Journal Articles

Evaluation of mechanical properties and nano-meso structures of 9-11%Cr ODS steels

Tanno, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Oba, Yojiro*; Onuma, Masato*; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Materials, 440(1-3), p.568 - 574, 2013/09

 Times Cited Count:10 Percentile:32.61(Materials Science, Multidisciplinary)

This study carried out mechanical tests and microstructure characterizations of several 9Cr and 11Cr-ODS tempered martensitic steels, and discussed the appropriate chemical composition range of 11Cr-ODS tempered martensitic steel from the viewpoint of high-temperature strength improvement. It was shown that the residual $$alpha$$-ferrite fraction in 11Cr-ODS steel was successfully controlled to the same level as the 9Cr-ODS steel by selecting the matrix chemical compositions on the basis of the multi-component phase diagram. The tensile strength decreased with decreasing W content from 2.0 to 1.4 wt%. On the other hand, creep strength at 973 K did not degrade by the decreasing W content. Both tensile strength and creep strength increased with increasing population of the nano-sized oxide particles. Small angle X-ray scattering analysis revealed that titanium and excess oxygen contents were key parameters in order to improve the dispersion condition of nano-sized oxide particles.

Journal Articles

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Science and Technology, 50(5), p.470 - 480, 2013/05

 Times Cited Count:3 Percentile:68.05(Nuclear Science & Technology)

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors such as microstructure instability and fuel pin rupture occurred. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change of 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

Journal Articles

ODS cladding fuel pins irradiation tests using the BOR-60 reactor

Kaito, Takeji; Yano, Yasuhide; Otsuka, Satoshi; Inoue, Masaki; Tanaka, Kenya; Fedoseev, A. E.*; Povstyanko, A. V.*; Novoselov, A.*

Journal of Nuclear Science and Technology, 50(4), p.387 - 399, 2013/04

 Times Cited Count:7 Percentile:41.73(Nuclear Science & Technology)

In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burnup of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed indicating good application prospects for high burnup fuel. On the other hand, peculiar irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the peculiar irradiation behaviors.

107 (Records 1-20 displayed on this page)