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Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)Amaya, Masaki
High Temperature Corrosion of Materials, 15 Pages, 2024/00
Times Cited Count:0 Percentile:0.02(Metallurgy & Metallurgical Engineering)Mohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*
Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11
Times Cited Count:0 Percentile:0(Materials Science, Multidisciplinary)Takeda, Takeshi
JAEA-Data/Code 2023-007, 72 Pages, 2023/07
An experiment denoted as IB-HL-01 was conducted on November 19, 2009 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment IB-HL-01 simulated a 17% hot leg intermediate break loss-of-coolant accident due to a double-ended guillotine break of pressurizer surge line in a pressurized water reactor (PWR). The break was simulated by a long nozzle upwardly mounted flush with a hot leg inner surface. The test assumptions included total failure of both high pressure injection system of emergency core cooling system (ECCS) and auxiliary feedwater system. In the experiment, relatively large size of break led to a fast transient of phenomena. The primary pressure steeply dropped after the break, and became lower than steam generator (SG) secondary-side pressure. Break flow turned from single-phase flow to two-phase flow soon after the break. Core uncovery started simultaneously with liquid level drop in downflow-side of crossover leg before loop seal clearing (LSC). The LSC was induced in both loops by steam condensation on accumulator (ACC) coolant of ECCS injected into cold legs. The whole core was quenched owing to the rapid recovery in the core liquid level after the LSC. Peak cladding temperature of simulated fuel rods was detected almost concurrently with the LSC. During the ACC coolant injection, liquid levels recovered in the hot legs and SG inlet plena because of liquid entrainment from the hot leg into the SG inlet plenum by high-velocity steam flow. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment IB-HL-01.
Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09
Takahatake, Yoko; Ambai, Hiromu; Sano, Yuichi; Takeuchi, Masayuki; Koizumi, Kenji; Sakamoto, Kan*; Yamashita, Shinichiro
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10
The corrosion behaviour of FeCrAl-ODS steels for the accident tolerant fuel cladding of LWRs were investigated in nitric acid solutions for the reprocessing process of spent fuels. The corrosion tests were carried out at 60C, 80C and the boiling point of the solutions, and the specimens were then analysed by XPS. The corrosion remarkably progressed at the boiling point, and the highest corrosion rate was 0.22 mm/y. In the oxide film, the atomic concentration of Fe was lower, than that in the base material, and those of Cr and Al were higher. The results show that the corrosion of FeCrAl-ODS steels in hot nitric acid solution is not severe because of the high corrosion resistance of the oxide film formed on the material; hence, the corrosion resistance of the new cladding materials in the dissolution process of spent fuel is acceptable for reprocessing operations.
Negyesi, M.; Amaya, Masaki
Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10
Times Cited Count:6 Percentile:50.9(Nuclear Science & Technology)Narukawa, Takafumi; Amaya, Masaki
Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11
Times Cited Count:10 Percentile:67.99(Nuclear Science & Technology)Udagawa, Yutaka; Nagase, Fumihisa; Fuketa, Toyoshi
JAERI-Research 2005-020, 40 Pages, 2005/09
In order to investigate effects of quenching temperature and cooling rate before quench on cladding ductility reduction under LOCA conditions, samples cut from non-irradiated 1717-type Zircaloy-4 cladding tubes for PWRs were oxidized in steam at 1373 and 1473 K, cooled at 2 to 7 K/s, and quenched at 1073 to 1373 K. The quenched samples were subjected to ring compression test, microstructure observation, and Vickers hardness test. Quenching temperature decrease obviously increased area fraction of phase in the radial cross section of the cladding, and reduced cladding ductility. Slow-cooling rate decrease increased unit size and hardness of precipitated phase, while phase area fraction and cladding ductility were not significantly changed. phase is harder than the surrounding region in the metallic layer and has higher oxygen content, indicating its low ductility. Consequently, increase in the area fraction in the cladding is a main cause of the reduction in cladding ductility with decrease in the quenching temperature.
Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi
Journal of Nuclear Science and Technology, 40(4), p.213 - 219, 2003/04
Times Cited Count:69 Percentile:96.6(Nuclear Science & Technology)Isothermal oxidation tests in flowing steam were performed on low-Sn Zircaloy-4 cladding tubes over the wide temperature range from 773 to 1573 K in order to obtain oxidation kinetics applicable to various loss-of-coolant accident conditions of LWRs. The oxidation generally obeys a parabolic rate law for the examined time range up to 3600s at temperatures from 1273 K to 1573K, and for a limited time range up to 900s from 773 to 1253 K. A cubic rate law is preferable for evaluating the longer-term oxidation at 1253 K and below. The parabolic rate law constant and the cubic rate law constant for measured weight gain were evaluated at every examined temperature, and Arrhenius-type equations were determined in order to describe the temperature dependence of the rate constants. It was indicated that the change of the oxidation kinetics from the cubic to the parabolic rate and the discontinuities in the temperature dependence of the rate constants are caused by the monoclinic/tetragonal phase transformation of ZrO.
Ogawa, Hiroaki*; Kiuchi, Kiyoshi
JAERI-Research 2002-037, 48 Pages, 2002/12
The difference in hydrogen permeation among candidate cladding materials such as 25Cr-35Ni stainless steel, Nb liner and reference materials such as 18Cr-8Ni SS, and Zr of Zircaloy base metal were evaluated by low energy plasma permeation simulated to hydrogen excited by heavy neutron irradiation. RF excitation source was arranged for the experimental apparatus in cooperating with temperature and bias control. Comparing with the thermodynamic gas driven permeation (GDP) in the same hydrogen pressure, the hydrogen permeation rate by the plasma driven permeation (PDP) was markedly accelerated at low to medium temperature range. The temperature dependency showed a knick at around 530K due to hydrogen-defect interactions. Comparing with Zr, Nb showed the high hydrogen solubility without the degradation by hydrate formation that is required to a getter material. The difference in PDP among candidates was analyzed with a new dissolution model for hydrogen.
Nagase, Fumihisa; Uetsuka, Hiroshi
JAERI-Research 2002-023, 23 Pages, 2002/11
To obtain basic data to evaluate fuel rod integrity during abnormal transient and accident of LWRs, high burnup PWR fuel claddings were heated for 0 to 600s at temperatures of 673 through 1173K, and the mechanical property changes were examined by using ring tensile test at room temperature. As a result of the test, it was shown that strength and ductility of the cladding are changed depending on heating temperature and time. The mechanical property changes by temperature transients are considered to be correspondent mainly to recovery of irradiation defect, recovery and recrystallization of the Zircaloy, phase transformations, and associated change of the hydride distribution and morphology. Comparison with unirradiated claddings suggested that irradiation effects are not completely annealed out by the short-term annealing at high temepratures. Radial change of hydrogen concentration was measured for the high burnup PWR fuel cladding and very high hydrogen concentration of about 2400wtppm was detected at the cladding periphery.
Kiuchi, Kiyoshi; Ioka, Ikuo; Tachibana, Katsumi; Suzuki, Tomio; Fukaya, Kiyoshi*; Inohara, Yasuto*; Kambara, Shozo; Kuroda, Yuji*; Miyamoto, Satoshi*; Ogura, Kazutomo*
JAERI-Research 2002-008, 63 Pages, 2002/03
no abstracts in English
Tanzawa, Sadamitsu; ;
Journal of Nuclear Science and Technology, 30(4), p.281 - 290, 1993/04
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)no abstracts in English
Tanzawa, Sadamitsu; Ishijima, Kiyomi
JAERI-M 91-183, 31 Pages, 1991/11
no abstracts in English
Nagase, Fumihisa; ; Uetsuka, Hiroshi; Furuta, Teruo
JAERI-M 90-165, 35 Pages, 1990/09
no abstracts in English
Hoshiya, Taiji; ; *;
JAERI-M 89-199, 35 Pages, 1989/12
no abstracts in English
; ; Tasaka, Kanji
Journal of Nuclear Science and Technology, 25(2), p.169 - 179, 1988/02
no abstracts in English
;
Journal of Nuclear Science and Technology, 21(7), p.515 - 527, 1984/00
Times Cited Count:10 Percentile:70.64(Nuclear Science & Technology)no abstracts in English
;
JAERI-M 83-068, 18 Pages, 1983/04
no abstracts in English