Tensile properties of irradiated modified 316 stainless steel (PNC316) at slow strain rates

Yano, Yasuhide; Miyazawa, Takeshi; Tanno, Takashi; Akasaka, Naoaki; Yoshitake, Tsunemitsu; Kaito, Takeji; Otsuka, Satoshi

Journal of Nuclear Science and Technology, 8 Pages, 2025/00

https://doi.org/10.1080/00223131.2025.2475164

The effects of strain rate on tensile properties of irradiated modified 316 stainless steel (PNC316) claddings were investigated. PNC316 claddings were irradiated at the experimental fast reactor Joyo using CRT402 control rod assembly at 400C up to 25 dpa. Post-irradiation ring tensile tests were carried out at strain rates of 3.310, 3.310 and 3.310 s at a test temperature of 350C. The results showed no obvious dependence of all strain rates on tensile properties, although a slight decrease in total elongation was observed at the slowest strain rate of 3.310 s. In addition, only a part of fracture surface at the slowest strain rate showed intergranular type region in the inner surface area, although the grain boundary separation occurred on inner surfaces near the fracture region at all strain rates. It is suggested that presence of a high content of helium near the inner surfaces would be related to the fracture behavior.

Oxide particles in oxide dispersion strengthened steel neutron-irradiated up to 158 dpa at Joyo

Toyama, Takeshi*; Tanno, Takashi; Yano, Yasuhide; Inoue, Koji*; Nagai, Yasuyoshi*; Otsuka, Satoshi; Miyazawa, Takeshi; Mitsuhara, Masatoshi*; Nakashima, Hideharu*; Onuma, Masato*; et al.

Journal of Nuclear Materials, 599, p.155252_1 - 155252_14, 2024/10

https://doi.org/10.1016/j.jnucmat.2024.155252

We investigated the stability of oxide nano particles in oxide dispersion-strengthened (ODS) steel, which is a promising candidate material for next-generation reactors, under neutron irradiation at high temperature to high doses. MA957, a 14Cr-ODS steel, was irradiated with Joyo in Japan Atomic Energy Agency under irradiation conditions of 130 dpa at 502C, 154 dpa at 589C, and 158 dpa at 709C. Three-dimensional atom probe (3D-AP) and transmission electron microscope (TEM) observation were performed to characterize the oxide particles in the ODS steels. A high number density of Y-Ti-O particle was observed in the unirradiated and irradiated samples. Almost no change in the morphology of the oxide particles, i.e. average diameter, number density, and chemical composition, has been observed in the samples irradiated to 130 dpa at 502C and to 154 dpa at 589C. A slight decrease in number density was observed in the sample irradiated to 158 dpa at 709CC. The hardness of any of the irradiated samples was almost unchanged from that of the unirradiated sample. It was revealed that the oxide particles existed stable, and the strength of the material was sufficiently maintained even after being neutron irradiated to high dose of 160 dpa at high temperature up to 700C. A part of this study includes the results of MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214482.

Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700C to 1000C

Imagawa, Yuya; Hashidate, Ryuta; Miyazawa, Takeshi; Onizawa, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.

Journal of Nuclear Science and Technology, 61(6), p.762 - 777, 2024/06

https://doi.org/10.1080/00223131.2023.2269178

The Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650C to 850C. However, little data have been obtained above 850C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700C to 1000C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix's phase transformation, and a single equation can express a creep rupture strength from 700C to 1000C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.

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, 61(4), p.521 - 529, 2024/04

https://doi.org/10.1080/00223131.2023.2243943

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 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 500C 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.

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

https://doi.org/10.1016/j.jnucmat.2022.154032

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 and 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.

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

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

https://doi.org/10.1016/j.jnucmat.2021.153105

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 600C 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 450C 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.

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

https://doi.org/10.1016/j.jnucmat.2021.152833

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 700C 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.

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.