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Journal Articles

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

Times Cited Count：3 Percentile：24.57(Nuclear Science & Technology)

JAEA Reports

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

Journal Articles

Effect of absorbed hydrogen on the stress corrosion cracking (SCC) susceptibility of unirradiated Zircaloy cladding

Amaya, Masaki; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(11), p.1091 - 1099, 2004/11

https://doi.org/10.3327/jnst.41.1091

Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)

Effect of absorbed hydrogen on the stress corrosion cracking (SCC) susceptibility of unirradiated Zircaloy cladding was examined. The data obtained from literatures show that the ratios of SCC threshold stress ( $_{th}$ ) to 0.2% yield stress ( $_{0.2}$ ) in unirradiated Zircaloy claddings increase with increasing hydrogen contents below 60 ppm, irrespective of the kind of Zircaloy-2 and -4. Thermodynamic calculations were carried out for the reaction between iodine gas and zirconium containing hydrogen. The results suggested that the reactions hardly occurred at increased hydrogen content and zirconium reacted with iodine gas only below 90 ppm of hydrogen. Since these tendencies correspond to those of the ratios of $_{th}$ to $_{0.2}$ on the hydrogen content, it is considered that hydrogen affects the reactions between iodine gas and zirconium and reduces the SCC susceptibility of Zircaloy claddings.

Journal Articles

Ion beam surface modification of Y-TZP and effects of subsequent annealing

Motohashi, Yoshinobu*; Shibata, Taiju; Harjo, S.*; Sakuma, Takaaki*; Ishihara, Masahiro; Baba, Shinichi; Sawa, Kazuhiro

Proceedings of 14th International Federation for Heat Treatment and Surface Engineering Congress Transactions of Materials and Heat Treatment Vol.25 No.5, p.1032 - 1036, 2004/10

no abstracts in English

Journal Articles

Radiation damage effects in superplastic 3Y-TZP irradiated with Zr ions

Motohashi, Yoshinobu*; Kobayashi, Tomokazu*; Harjo, S.*; Sakuma, Takaaki*; Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hoshiya, Taiji

Nuclear Instruments and Methods in Physics Research B, 206, p.144 - 147, 2003/05

https://doi.org/10.1016/S0168-583X(03)00702-X

Times Cited Count：4 Percentile：33.23(Instruments & Instrumentation)

The 3mol% yttria containing tetragonal zirconia polycrystals(3Y-TZP) were irradiated using 130MeV Zr $^{11+}$ ions in the TANDEM accelerator facility at Tokai Research Establishment, JAERI. Irradiation was performed with the fluence of 3.510 $^{16}$ and 2.110 $^{17}$ ions/m $^{2}$ . Residual stress and changes in mechanical properties caused by the ion irradiation and the effects of the subsequent annealing are studied. The occurrence of compressive residual stresses and increases in hardness and fracture toughness were found at the surface regions of as-irradiated specimens. It was found from the subsequent annealing that these quantities decreased gradually with raising the annealing temperature and returned to that of un-irradiated state at around 1173K. A most probable cause of the increases in the hardness and fracture toughness after the irradiation may, therefore, be the residual compressive stresses left in the irradiated surface region.