Corrosion behavior and mechanical properties of Modified SUS316 fuel cladding for fast reactors under a high-temperature sodium environment

Imagawa, Yuya; Toyota, Kodai; Onizawa, Takashi; Kato, Shoichi

JAEA-Data/Code 2024-010, 90 Pages, 2024/11

JAEA-Data-Code-2024-010.pdf:5.41MB

To establish a material testing technique in sodium and to develop a method to evaluate the sodium environmental effects, sodium tests on fast reactor fuel cladding have been carried out. Fast reactor fuel cladding is susceptible to corrosion thinning and compositional change due to sodium because of its high temperature (around 675C) and thin wall (about 0.5 mm) during normal operation. Therefore, it is important to evaluate the corrosion behavior and mechanical properties under a high-temperature sodium environment. This report summarizes the results of experimental studies on corrosion behavior and mechanical properties of modified type 316 stainless steel fuel cladding applied to actual fast reactors under a high-temperature sodium environment, in order to reflect the results to future research activities and to consolidate knowledge and experience.

Validation of the applicability of the best-fit fatigue curves for 550C in Mod.9Cr-1Mo steel to 110 cycles

Toyota, Kodai; Imagawa, Yuya; Onizawa, Takashi; Kato, Shoichi; Furuya, Yoshiyuki*

Nihon Kikai Gakkai Rombunshu (Internet), 89(928), p.23-00206_1 - 23-00206_15, 2023/12

https://doi.org/10.1299/transjsme.23-00206

In order to design fast reactors, it is necessary to consider high cycle fatigue of structural materials up to 110 cycles; to evaluate high cycle fatigue at 110 cycles, it is necessary to develop a best-fit fatigue curve applicable up to 110 cycles. In this study, high cycle fatigue tests were conducted under strain-controlled conditions and ultrasonic fatigue tests were also conducted to develop a high cycle fatigue evaluation method for Mod.9Cr-1Mo steel, which is a candidate material for fast reactor structural materials. Based on the test results, the best-fit fatigue curves were extended and the applicability of the JSME best-fit fatigue curves up to 110 cycles was verified.

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.

Proposal of maintenance rationalization for next-generation fast reactors based on the analysis of the prolonged maintenance of the prototype fast-breeder reactor in Japan, "Monju", 2; Analysis of maintenance plan of "Monju" in reactor cold shutdown

Toyota, Kodai; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru

Hozengaku, 20(2), p.95 - 103, 2021/07

Proposal of maintenance rationalization for next-generation fast reactors based on the analysis of the prolonged maintenance of the prototype fast-breeder reactor in Japan, "Monju", 1; Analysis of plant schedule of "Monju" in cold shutdown

Hashidate, Ryuta; Toyota, Kodai; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru

Hozengaku, 19(4), p.115 - 122, 2021/01

In order to improve both safety and economic efficiency of a nuclear power plant, it is necessary to realize rational maintenance based on characteristics of the plant. The prototype fast-breeder reactor in Japan, Monju, spent most of the year for the maintenance. Thus, it is important to identify causes of the prolonged maintenance of Monju and to investigate countermeasures for implementation of rational maintenance of next-generation fast reactors. In this study, the authors investigated the causes of the prolonged maintenance of Monju during reactor cold shutdown based on the plant schedule of Monju. In addition, we proposed the maintenance optimization idea for next-generation fast reactors to solve the revealed issues.

Characterization of iron-based and tungsten-based high-entropy alloys

Wakai, Eiichi; Noto, Hiroyuki*; Shibayama, Tamaki*; Iwamoto, Yosuke; Sato, Koichi*; Yano, Yukihiro*; Yoshikawa, Maya*; Nakagawa, Yuki*; Toyota, Kodai; Onizawa, Takashi

no journal, , 

In this study, we attempted to fabricate Fe-based and W-based Fe-HEA and W-HEA, both of which have been studied extensively in recent years to enhance the durability of equipment used in high-energy beam irradiation environments. For Fe-based HEAs, Fe-Mn-Cr-V-Al-C alloys were melted and casted, and then subjected to homogenization heat treatment (homogenization heat treatment (1150C for 2h)). After homogenization heat treatment, a 3-point bending test was performed at room temperature. The homogenization heat treatment resulted in an increase in ductility in the 3-point bending test and a decrease in elastic modulus based on ultrasonic velocity measurements. XRD measurements of this material after heat treatment (800C for 10 min and Water Quenched) showed that it has a BCC structure and a Vickers hardness that exceeds that of pure W. On the other hand, in the preparation of W-based HEA material (W-Fe-Si-V-Cr alloy), an arc melting method using powder was attempted, and it was found that an almost homogeneous crystallized alloy could be produced.

The Material property equations for 316FR steel at extremely high temperature

Okuda, Takahiro; Yamashita, Hayato; Toyota, Kodai; Shimomura, Kenta; Onizawa, Takashi; Kato, Shoichi

no journal, , 

This study describes the setting of the material property equations of 316FR steel at an extremely high temperature which can be applied to severe accident conditions of generation IV fast reactors. 316FR steel will be applied to structural materials, e.g. reactor vessel, in the generation IV fast reactors. After the severe accident in Fukushima Daiichi Nuclear Power Plants, the evaluation of structural integrity was found to be very important severe accident condition. The development of the generation IV fast reactors requires the material properties of 316FR steel at the extremely high temperature. However, such data has not been acquired. Therefore, tensile and creep tests were carried out in the temperature range over 700C for 316FR steel. Based on the acquired data from the tests, the equations that can evaluate the material properties of 316FR steel at the extremely high temperature were set up. They are an elasto-plastic stress-strain equation, a creep rupture equation and a creep strain equation.

Development of high cycle fatigue evaluation method for fast reactor structural materials

Toyota, Kodai

no journal, , 

no abstracts in English

Evaluation of defect number density in sodium equipment stainless steel welds

Toyota, Kodai; Hashidate, Ryuta; Yada, Hiroki; Takaya, Shigeru; Miyakoshi, Hiroyuki; Kato, Shoichi

no journal, , 

Probabilistic fracture mechanics (PFM) evaluation requires information on the probabilistic distributions of the number and size of initial defects, material properties such as crack growth rate due to fatigue and creep, and load to evaluate the failure probability of components. In this study, ultrasonic testing was conducted on the welds of the test equipment used in the research and development of fast reactors, and the number and size of defects were evaluated. The results could be used as conservative values of initial defects, and the values related to initial defects for PFM evaluation of FBR components were examined.

Development of location identification method of reactor internal structures using ultrasonic phased array

Asahi, Manabu*; Tomita, Naoki*; Furuya, Masahiro*; Morita, Hidetoshi*; Toyota, Kodai

no journal, , 

To inspect reactor internal structures of Sodium-cooled Fast Reactors (SFRs) from the outer surface of vessel, the measurement method was proposed and developed to analyze noisy echo signals of an ultrasonic phased array flaw detector. The proposed method successfully reconstructs images of the stainless-steel structure by optimizing the measurement parameters.