Fatigue crack propagation experimental evaluation and modeling in an austenitic steel elbow from a LMFBR primary system piping

Garcia Rodriguez, D.; Sakakibara, Yasuhide*

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07

The behavior of high temperature (400C) low-cycle fatigue crack propagation during in-plane bending of an elbow from Monju LMFBR is presented in three stages. 1st, experimental measurements made in a real-size laboratory specimen are presented. This test was carried out under displacement-controlled conditions, with artificial defects introduced in the crown parts, where the maximum stress arises. 2nd, the experimental setup is simulated making use of FEA, in order to obtain the actual stress distribution through the loaded elbow. Finally, based on the FEA data, deterministic fatigue crack propagation based on the J-integral criterion is compared with the experimental data.

Influence of thermal aging on cast stainless steels used in JAEA's nuclear reactor Fugen

Nogiwa, Kimihiro; Onitsuka, Takashi; Abe, Teruyoshi; Sakakibara, Yasuhide; Horie, Kaoru; Nakamura, Takahisa

Journal of Nuclear Science and Technology, 50(9), p.883 - 890, 2013/09

https://doi.org/10.1080/00223131.2013.812511

The degree of influence of thermal aging on cast stainless steels over a protracted time at low temperatures was investigated by means of a toughness test and microstructural characterization performed on components dismantled from the advanced thermal proto-type reactor "Fugen". The thermal embrittlement mechanism was examined using data obtained dismantled from the materials that had aged on site. The results of a Charpy impact test and microstructural characterization performed using 3DAP analysis reveal early signs of a thermal aging effect over a protracted period at low temperatures corresponding boiling-water reactor (BWR).

Measurement and analysis of Echizen pottery anagama; Toward a scientific elucidation of the firing process

Terauchi, Makoto; Watanabe, Toshio; Sakakibara, Yasuhide; Tanaka, Teruhisa*; Masudaya, Kosei*

Shiseki "Suzu Tokigama Ato" Kuni Shitei Kinen Shimpojiumu Hokokusho, p.85 - 112, 2010/03

JAEA has had many kinds of meetings for technology exchange willingly as an activity to widely make use of the accumulation of the results of research and development on the way to develop FBR "Monju" in general industry. In this report, one of the local traditional industries of Fukui Prefecture, "Echizen pottery" is featured to show an activity of technological exchange. JAEA established a "technological exchange group on Echizen pottery" to scientifically find out the process until firing ends by analyzing behaviors including temperature distribution and flow condition of the inside of anagama with an analyzing technology (thermal Hydraulics), one of our representative research results. This is a report to show the content which has been discussed and evaluated based on the results of firing experiments and their analysis in this group.

Proceedings 4th Tsuruga International Energy Forum

Sakakibara, Yasuhide; IKEDA, Hiroshi; Yokota, Yoshio; Kitabata, Takuya; Yamaguchi, Toshihiko

JNC TN4200 2004-001, 317 Pages, 2004/08

JNC-TN4200-2004-001.pdf:24.97MB

None

None

Sakakibara, Yasuhide; ;

Saikuru Kiko Giho, (15), p.145 - 154, 2002/06

https://rdreview.jaea.go.jp/gihou/pdf2/n15-16.pdf

None

Training report of the FBR cycle comprehensive training facility in 2000

Sawada, Makoto; Watanabe, Tomoo; Iwasaki, Ryoji*; Sakakibara, Yasuhide; Nagai, Fumio*

JNC TN4410 2001-006, 53 Pages, 2001/06

JNC-TN4410-2001-006.pdf:9.03MB

The FBR Cycle Comprehensive Training Facility constructed in the International Cooperation and Technology Development Center (in Shiraki area) was opened in earnest since October in 2000. The facility is mainly with the aim of encouraging in technical skills of the operators and maintenance engineers in the prototype fast breeder rector Monju and also is offered to another center in JNC and the local community as a global training facility. Its facility consists of the called sodium school and maintenance school. It is able to study widely sodium huddling technologies, such as chemical and physical properties of sodium, sodium loop operating skill, sodium fire extinguish and treatment sodium compound, and counter training for sodium piping leakage at the sodium school. While the nine study models are installed in the maintenance school and the trainees can learn not only related Monju inherent maintenance technology but also some general maintenance skills. During for six months, October 2000 to March 2001, the six sodium training courses were held 22 times and the eight maintenance courses were carried out 11 times. Total participated number is 305 and 75 trainees among of them joined from the community. To do the training effectively, as the evaluation and analysis system of training result, the skill confirmation test of which aims to have a grasp training results quantitatively by themselves was devised and was introduced. Almost of the training course's results were improved remarkably such as that average point after getting training were risen level of 80 point from level of 40 point before training. Like this, it is obvious that their skills of the Monju operators and maintenance engineers are advancing toward to Monju restarting.

Large-scaled thermohydraulic tests plan for cooling systems in fast reactors; Experimental models of reactor vessel and the primary cooling system

Kamide, Hideki; Hayashi, Kenji; Gunji, Minoru; Hayashida, Hitoshi; Nishimura, Motohiko; Iitsuka, Toru; Kimura, Nobuyuki; Tanaka, Masaaki; Nakai, Satoru; Mochizuki, Hiroyasu; et al.

PNC TN9410 96-279, 51 Pages, 1996/08

PNC-TN9410-96-279.pdf:2.92MB

Large-scaled thermohydraulic tests are planned for some new key technologies in the heat transport systems of demonstration fast reactors, in which the reactor vessel, the primary system, the secondary system, water-steam system, and the decay heat removal systems are modeled. Thermohydraulic issues and structural integrity issues were discussed for the top entry piping systems with satellite pools of the intermediate heat exchangers and the pumps, the natural circulation decay heat removal using direct heat exchangers in a reactor hot pool, the reactor vessel wall cooling system, and the new type of steam generators in the demonstration reactor. Concepts of the experimental model for the reactor vessel and the primary system were created and compared with each other for the sodium test facility which enables to answer the thermohydraulic and structural integrity issues. Following items were considered in the creation and in the selection of the models; (1)solution of the issues for Demonstration First Reactor on total system characteristics, the reactor vessel wall cooling system, the decay heat removal system, and the steam generator, (2)balance between the thermohydraulic issues and the structural integrity issues, (3)simulations of compound phenomena and interactions between the components and the heat transport systems. Total system of test facility was specified based on the selected test model.

Study of thermal ageing behaviour of duplex stainless steels by positron annihilation lifetime spectroscopy

Onitsuka, Takashi; Nogiwa, Kimihiro; Abe, Teruyoshi; Sakakibara, Yasuhide; Nakamura, Takahisa; Horie, Kaoru

no journal, , 

no abstracts in English

Research activity of aging of nuclear power plants in JAEA

Sakakibara, Yasuhide

no journal, , 

This report describes the results of the research conducted by the JNC entrusted by JNES. It was aimed to improve the safety and the reliability of aged LWRs through the studies valuable for the regulatory activities. It was also utilizing the technologies developed at the Monju FBR and the materials in the Fugen NPS facilities.

Study of thermal ageing behaviour of duplex stainless steels by positron annihilation lifetime technique

Onitsuka, Takashi; Nogiwa, Kimihiro; Abe, Teruyoshi; Sakakibara, Yasuhide; Nakamura, Takahisa; Horie, Kaoru

no journal, , 

no abstracts in English