Oxidation and embrittlement behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki

Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12

https://doi.org/10.1016/j.jnucmat.2023.154736

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki

Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08

https://doi.org/10.1016/j.jnucmat.2023.154467

Hierarchical Bayes model to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under LOCA conditions

Narukawa, Takafumi; Hamaguchi, Shusuke*; Takata, Takashi*; Udagawa, Yutaka

Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

https://doi.org/10.1080/00223131.2020.1724207

Behavior of high-burnup advanced LWR fuel cladding tubes under LOCA conditions

Narukawa, Takafumi; Amaya, Masaki

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09

Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

Narukawa, Takafumi; Yamaguchi, Akira*; Jang, S.*; Amaya, Masaki

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

Development of core and structural materials for fast reactors

Asayama, Tai; Otsuka, Satoshi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06

This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design.

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

https://doi.org/10.1080/00223131.2015.1025884

Mechanical characteristics evaluation of fuel cladding tube for reduced-moderation water reactor, 1 (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Research 2005-005, 23 Pages, 2005/03

JAERI-Research-2005-005.pdf:1.65MB

Fuel elements used in The Reduced-Moderation Water Reactor (RMWR) have the lamellar structure consisting of MOX pellets and UO blankets in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod. Therefore, it is important to evaluate the local deformation behavior due to the transient temperature distribution. To estimate the thermal deformation behavior, the temperature and stress distribution of the fuel cladding tube assumed in the designed reactor were analyzed. Moreover, basic physical properties and mechanical properties for analyzing the deformation behavior were obtained by experiment using fuel cladding tubes made of candidate alloys. In addition, the appropriate experimental conditions for realizing the practical thermal deformation behavior of the fuel cladding tube was selected by adjusting the testing temperature distribution based on data obtained with thermal analysis.

Development of testing techniques to evaluate thermal deformation behavior of fuel cladding tubes (Contract research)

Kaneko, Tetsuji; Tsukatani, Ichiro; Kiuchi, Kiyoshi

JAERI-Tech 2004-035, 18 Pages, 2004/03

JAERI-Tech-2004-035.pdf:0.81MB

Fuel elements used in the Reduced-Moderation Water Reactor (RMWR) have the stacking structure consisting of MOX pellets and UO blankets in a fuel rod in order to attain the high breeding ratio and high burn-up simultaneously. It is a characteristic of the fuel elements that there is high thermal stress caused by inhomogeneous linear power density along the longitudinal direction of the fuel rod in comparison with the present LWR fuels. For this reason, it is important to estimate local deformation behavior of the fuel cladding tube with temperature difference caused by MOX pellet and UO blanket. The testing machine was designed to investigate thermal-fatigue behavior under biaxial stress condition. The testing machine consists of the temperature distribution control unit, low cycle fatigue testing unit and internal pressure loading unit, it is also possible to conduct the simulation tests to investigate effects of pressure change with burn-up and longitudinal load change due to operation modes and restriction of fuel rods.