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

Double diffusive dissolution model of UO pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 10 Pages, 2022/00

https://doi.org/10.1080/00223131.2022.2114956

Toward a more reliable evaluation of fuel fracture during LOCA; Development of uncertainty quantification methodology and evaluation of the effect of fuel burnup extension

Narukawa, Takafumi

Nihon Genshiryoku Gakkai-Shi ATOMO, 63(11), p.780 - 785, 2021/11

https://doi.org/10.3327/jaesjb.63.11_780

no abstracts in English

Steam oxidation of silicon carbide at high temperatures for the application as accident tolerant fuel cladding, an overview

Pham, V. H.; Kurata, Masaki; Steinbrueck, M.*

Thermo (Internet), 1(2), p.151 - 167, 2021/09

https://doi.org/10.3390/thermo1020011

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

https://doi.org/10.1016/B978-0-12-803581-8.11682-0

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

https://doi.org/10.1080/00223131.2019.1709996

The Effect of base irradiation on failure behaviors of UO and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

https://doi.org/10.1016/j.anucene.2019.107268

Microstructure and texture evolution and ring-tensile properties of recrystallized FeCrAl ODS cladding tubes

Aghamiri, S. M. S.*; Sowa, Takashi*; Ukai, Shigeharu*; Ono, Naoko*; Sakamoto, Kan*; Yamashita, Shinichiro

Materials Science & Engineering A, 771, p.138636_1 - 138636_12, 2020/01

https://doi.org/10.1016/j.msea.2019.138636

Oxide dispersion strengthened (ODS) FeCrAl ferritic steels are being developed as potential accident tolerance fuel cladding materials for the light water reactors (LWRs) due to significant improvement in steam oxidation by alumina forming scale and good mechanical properties up to high temperatures. In this study, the microstructural characteristics and tensile properties of the two FeCrAl ODS cladding tubes with different extrusion temperatures of 1100C and 1150C were investigated during processing conditions. While the hot extruded sample showed micron sized elongated grains with strong -fiber in 110 texture, cold pilger rolling process change the microstructure to submicron/micron size grain structure along with texture evolution to both -fiber (110 texture) and -fiber ({111} texture) via crystalline rotations. Subsequently, final annealing resulted in evolution of microstructure to large grain recrystallized structure starting at recrystallization temperature of 810-850C. Two distinct texture development happened in recrystallized cladding tubes, i.e., only large elongated grains of (110) 211 texture following extrusion temperature of 1100C; and two texture components of (110) 211 and {111} 112 following higher extrusion temperature of 1150C. The different texture development and retarding of recrystallization progress in 1100C-extruded cladding tubes were attributed to higher distribution of oxide particles.

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

Overview of accident-tolerant fuel R&D program in Japan

Yamashita, Shinichiro; Ioka, Ikuo; Nemoto, Yoshiyuki; Kawanishi, Tomohiro; Kurata, Masaki; Kaji, Yoshiyuki; Fukahori, Tokio; Nozawa, Takashi*; Sato, Daiki*; Murakami, Nozomu*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09

After the nuclear accident at Fukushima Daiichi Power Plant, research and development (R&D) program for establishing technical basis of accident-tolerant fuel (ATF) started from 2015 in Japan. Since then, both experimental and analytical studies necessary for designing a new light water reactor (LWR) core with ATF candidate materials are being conducted within the Japanese ATF R&D Consortium for implementing ATF to the existing LWRs, accompanying with various technological developments required. Until now, we have accumulated experimental data of the candidate materials by out-of-pile tests, developed fuel evaluation codes to apply to the ATF candidate materials, and evaluated fuel behavior simulating operational and accidental conditions by the developed codes. In this paper, the R&D progresses of the ATF candidate materials considered in Japan are reviewed based on the information available such as proceedings of international conference and academic papers, providing an overview of ATF program in Japan.