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

https://doi.org/10.1016/j.jnucmat.2019.01.052

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.

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

https://doi.org/10.1016/j.jnucmat.2018.03.054

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.

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.

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

https://doi.org/10.1016/j.nme.2016.10.007

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

https://doi.org/10.1016/j.nme.2016.02.009

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 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.

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

https://doi.org/10.1080/00223131.2014.964789

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 690C 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.

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 835C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

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

https://doi.org/10.1016/j.jnucmat.2012.10.038

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.

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

https://doi.org/10.1016/j.jnucmat.2013.04.006

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 -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.

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

https://doi.org/10.1080/00223131.2013.773165

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.