Effect of repair welding location and number of welds performed on creep strength of Mod. 9Cr-1Mo steel welded joints

Yamashita, Hayato; Toyota, Kodai; Onizawa, Takashi; Yamamoto, Kenji*; Kubo, Koji*

Nihon Kikai Gakkai Rombunshu (Internet), 92(955), p.25-00176_1 - 25-00176_13, 2026/03

https://doi.org/10.1299/transjsme.25-00176

It is planned that Mod.9Cr-1Mo steel will be used as the structural material for the steam generator of the demonstration fast reactor. Creep strength of welded joints of Mod.9Cr-1Mo steel is lower than that of base metal at high temperature and long time. In addition, the creep strength of welded joints of Mod.9Cr-1Mo steel with repair welding is lower than that of without repair welding. In this study, the effects of the location and number of repair welds performed on the creep strength of welded joints of Mod.9Cr-1Mo steel were investigated, and a repair welding method was developed for the construction of the demonstration fast reactor. It was clarified that, regardless of the location and number of repair welds performed, the degradation in creep strength could be greatly reduced by performing the repair welding before post weld heat treatment (PWHT). Furthermore, it was found that the superposition of thermal history and the formation of a heat affected zone (HAZ) on the weld metal promote the coarsening of ferrite grains during creep and cause a slight reduction in creep strength. All repair welding was performed prior to PWHT, and a repair welding method was also developed to minimize thermal history superposition and weld metal HAZ formation.

Development of a new materials surveillance technology for fast reactors

Okajima, Satoshi; Ando, Masanori; Toyota, Kodai; Ishigami, Katsuo*; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai; Wakai, Takashi

Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 10 Pages, 2025/07

https://doi.org/10.1115/PVP2025-153438

This paper describes current efforts within the JAEA to develop a new materials surveillance technology for SFRs that aims at raising the confidence levels in operating reactors' structural integrity to the ones which could otherwise be achieved only by a vast amount of long-term material test data that needs decades to accumulate. The technology would allow for monitoring on-going materials degradation processes in an actual reactor, namely fatigue and creep-fatigue, by inserting a newly developed specimen for this purpose into the reactor and establishing procedures to observe and evaluate the specimen to get desired information such as the extent of degradation and residual life. To establish the test procedure, a passive creep-fatigue test model was developed in the framework of CNWG, a collaboration agreement between the US and Japan, and a demonstration test was performed under cyclic thermal loads using an electric furnace. As a result, the test specimen failed by several hundred thermal cycles. Macroscopic inspection of the specimen after the test showed the test had been performed successfully without buckling. Fracture surface observations suggested that the failure caused probably by fatigue or creep-fatigue.

Evaluation of stability of precipitates under irradiation in 316FR steel used as fast reactor structural material

Toyota, Kodai; Onizawa, Takashi; Wakai, Eiichi*

Research & Development in Material Science (Internet), 21(5), p.2632 - 2637, 2025/06

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.

Study of multi-simultaneous measurement system development under radiation environment and innovative materials

Wakai, Eiichi; Iwamoto, Yosuke; Shibayama, Tamaki*; Sato, Koichi*; Toyota, Kodai; Onizawa, Takashi; Wakui, Takashi; Ishida, Taku*; Makimura, Shunsuke*; Nakagawa, Yuki*; et al.

no journal, , 

In the fields of accelerator target systems, nuclear power, aerospace, etc., radiation degradation of structural materials and equipment occurs, and therefore, the development of materials with high durability and excellent functions is expected. In order to create innovative materials that can be used in radiation fields, we are developing a new non-destructive inspection technique that can accurately measure the internal defects of various materials in radiation fields. As an innovative material, high-entropy alloys (HEA) are known for their high strength and ductility, and are expected to be used in various applications. In this talk, we will report on the construction of a measurement principle that enables multi-simultaneous measurements even in radiation fields, the status of HEA prototypes, and the status and progress of irradiation analysis of metals and other materials.

Development of the material strength standard of 316FR steel and modified 9Cr-1Mo steel for next-generation fast reactor in Japan

Onizawa, Takashi; Toyota, Kodai; Imagawa, Yuya; Okajima, Satoshi; Ando, Masanori

no journal, , 

In order to realize a fast reactor that achieves both safety and economic efficiency at a high level, Japan Atomic Energy Agency (JAEA) is developing the material strength standard for fast reactor design. JAEA has developed the material strength standard based on the acquired data and its evaluation results, and the standard have been incorporated in the Japan Society of Mechanical Engineers (JSME) code, Rules on the Design and Construction of Nuclear Power Plants, Section II, Fast Reactors (JSME D&C FRs Code). This paper describes the standard that recently incorporated in the JSME D&C FRs code and ongoing studies for improvements in the near future.