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Journal Articles

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

Hydrogen absorption behavior on zirconium under $$gamma$$-radiolysis of nitric acid solution

Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi

Nihon Genshiryoku Gakkai Wabun Rombunshi, 16(2), p.100 - 106, 2017/05

Zirconium (Zr) has been used as a structural material at the spent nuclear fuel reprocessing plant in Japan because of its excellent corrosion resistance against nitric acid solution. And the radiolytic hydrogen is known to be generated in the spent nuclear fuel solution. Zr is known to be highly susceptible to hydrogen embrittlement. Therefore, evaluating the radiolytic hydrogen absorption behavior of Zr in nitric acid solution (HNO$$_{3}$$) is essential. In this study, immersion tests were conducted on Zr in nitric acid solutions under $$gamma$$-ray irradiation to evaluate its radiolytic hydrogen absorption behavior. Results showed that hydrogen concentration on Zr increased both in 1-3 mol/L HNO$$_{3}$$ and pure water at 5 and 7 kGy/h after immersion. The amount of hydrogen absorption on Zr under $$gamma$$-ray irradiation had a direct correlation with the radiolytic hydrogen generation value in HNO$$_{3}$$. The results of glow discharge optical emission spectrometry, thermal desorption spectroscopy, and X-ray diffraction result shows that the absorbed radiolytic hydrogen generated a hydride on the surface of Zr.

Journal Articles

Embrittlement and fracture behavior of pre-hydrided cladding under LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.668 - 677, 2005/10

A systematic research program on high burnup fuel behavior under LOCA conditions is being conducted at JAERI. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence were conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44 GWd/t at a PWR, to investigate behavior and condition of cladding fracture during quenching for safety evaluation. Differences were not clearly observed between irradiated and unirradiated claddings at similar hydrogen concentrations in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases. Ductility of pre-hydrided, oxidized and quenched claddings was also evaluated by using ring-tensile and ring-compression tests. Embrittlement criteria (zero-ductility limits) from both the tests were lower than the fracture conditions in the integral thermal shock tests. This indicates that loading conditions should be well simulated to evaluate cladding performance under LOCA conditions.

Journal Articles

Suppression of hydrogen absorption to V-4Cr-4Ti alloy by TiO$$_{2}$$/TiC coating

Hirohata, Yuko*; Motojima, Dai*; Hino, Tomoaki*; Sengoku, Seio

Journal of Nuclear Materials, 313-316(1-3), p.172 - 176, 2003/03

 Times Cited Count:12 Percentile:63.39(Materials Science, Multidisciplinary)

Titanium oxide film was coated on entire surface of a V-4Cr-4Ti alloy to reduce hydrogen absorption at low temperature region. The film consisted with a mixture of TiO2 and TiC. The content of TiO$$_{2}$$ was approximately 80%. The hydrogen absorption rate of Ti-oxide coated V-alloy largely decreased with increase of the film thickness. In the case of the film with a thickness of 0.5 micron, the absorption rate was fifty times smaller than that of non-coated V-alloy at the absorption temperature of 573 K.

Journal Articles

Positron annihilation studies of defects in ion implanted palladium

Abe, Hiroshi; Uedono, Akira*; Uchida, Hirohisa*; Komatsu, A.*; Okada, Sohei; Ito, Hisayoshi

Materials Science Forum, 363-365, p.156 - 158, 2001/05

no abstracts in English

Journal Articles

Estimation of conservatism of present embrittlement criteria for zircaloy fuel cladding under LOCA

; ;

Zirconium in the Nuclear Industry, p.734 - 746, 1984/00

no abstracts in English

JAEA Reports

Analysis of Behavior on Solution and Diffusion of Hydrogen in Iron

JAERI-M 83-052, 66 Pages, 1983/03

JAERI-M-83-052.pdf:1.84MB

no abstracts in English

Journal Articles

Reaction behavior of zircaloy-4 in steam-hydrogen mixture at high temperature

;

Journal of Nuclear Materials, 105(2), p.119 - 131, 1982/00

 Times Cited Count:19 Percentile:85.33(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Embrittlement of Zircaloy-4 due to oxidation in environment of stagnant steam

; ;

Journal of Nuclear Science and Technology, 19(2), p.158 - 165, 1982/00

 Times Cited Count:19 Percentile:85.33(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Zircaloy-4 cladding embrittlement due to inner surface oxidation under simulated LOCA condition

; ;

Journal of Nuclear Science and Technology, 18(9), p.705 - 717, 1981/00

 Times Cited Count:39 Percentile:95.72(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Hydrogen Absorption by Zircaloy Cladding being Caused by Inner-Surface Oxidation

; ; ; ;

JAERI-M 8497, 27 Pages, 1979/10

JAERI-M-8497.pdf:1.1MB

no abstracts in English

Oral presentation

Study on tritium confinement method using Li rod with Zr in very high temperature gas-cooled reactor; Hydrogen storage properties of Zr in high temperature (700$$sim$$850$$^{circ}$$C) conditions

Okamoto, Ryo*; Matsuura, Hideaki*; Ida, Yuma*; Koga, Yuki*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; Nagasumi, Satoru; et al.

no journal, , 

Currently, many researches to achieve DT nuclear-fusion power generation are under proceeding but the method to provide initial tritium loaded to fusion prototype reactor is not clear. The method of tritium production by using high temperature gas-cooled reactor (HTGR) was proposed. In this method, lithium rods are loaded to the reactor core of HTGR and tritium is produced by $$^{6}$$Li(n,$$alpha$$)T reaction. And the method to reduce the spilled tritium by using the lithium rod with zirconium layer was proposed. In this study, the experiments to evaluate the performance of hydrogen absorption in the zirconium layer were conducted under the temperature condition more than 700$$^{circ}$$C which is the normal operation condition for the very high temperature gas-cooled reactor (VHTR). The experimental result concerning solubility and diffusion factor of hydrogen in the zirconium layer will be presented and discussed.

Oral presentation

Study on T-production Li rod for high temperature gas cooled reactor; Temperature dependence of the hydrogen absorption speed in Zr

Nakagawa, Kyoichi*; Matsuura, Hideaki*; Koga, Yuki*; Katayama, Kazunari*; Otsuka, Teppei*; Goto, Minoru; Hamamoto, Shimpei; Ishitsuka, Etsuo; Nakagawa, Shigeaki; Tobita, Kenji*; et al.

no journal, , 

The containment of tritium by using zirconium in the lithium rod applied to the tritium production in the HTGR has been investigated. Considering the temperature distribution on the reactor core, the containment performance to tritium was evaluated through the measurement of the Zr hydrogen absorption rate in its temperature range.

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