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Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06
Times Cited Count：5 Percentile：72.25(Nuclear Science & Technology)
-radiolysis of nitric acid solutionIshijima, 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
) is essential. In this study, immersion tests were conducted on Zr in nitric acid solutions under -ray irradiation to evaluate its radiolytic hydrogen absorption behavior. Results showed that hydrogen concentration on Zr increased both in 1-3 mol/L HNO and pure water at 5 and 7 kGy/h after immersion. The amount of hydrogen absorption on Zr under -ray irradiation had a direct correlation with the radiolytic hydrogen generation value in HNO. 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.
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.
/TiC coatingHirohata, Yuko*; Motojima, Dai*; Hino, Tomoaki*; Sengoku, Seio
Journal of Nuclear Materials, 313-316(1-3), p.172 - 176, 2003/03
Times Cited Count：13 Percentile：64.08(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 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.
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
; ;
Zirconium in the Nuclear Industry, p.734 - 746, 1984/00
no abstracts in English
JAERI-M 83-052, 66 Pages, 1983/03
no abstracts in English
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Journal of Nuclear Materials, 105(2), p.119 - 131, 1982/00
Times Cited Count：20 Percentile：85.73(Materials Science, Multidisciplinary)no abstracts in English
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Journal of Nuclear Science and Technology, 19(2), p.158 - 165, 1982/00
Times Cited Count：24 Percentile：88.62(Nuclear Science & Technology)no abstracts in English
-Steam Reaction and Rapid Cooling; ;
JAERI-M 9681, 19 Pages, 1981/09
no abstracts in English
; ;
JAERI-M 9445, 37 Pages, 1981/04
no abstracts in English
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Journal of Nuclear Science and Technology, 18(9), p.705 - 717, 1981/00
Times Cited Count：40 Percentile：95.62(Nuclear Science & Technology)no abstracts in English
; ; ; ;
JAERI-M 8497, 27 Pages, 1979/10
no abstracts in English
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.
850
C) conditionsOkamoto, 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 
Li(n,
)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 700C 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.
