Shimizu, Kazuyuki*; Toda, Hiroyuki*; Fujihara, Hiro*; Yamaguchi, Masatake; Uesugi, Masayuki*; Takeuchi, Akihisa*; Nishijima, Masahiko*; Kamada, Yasuhiro*
Corrosion, 79(8), p.818 - 830, 2023/08
7xxx aluminum alloys are representative high-strength aluminum alloys; however, mechanical property degradation due to hydrogen hinders further strengthening. We propose the dispersion of Mn-based second-phase particles as a novel technique for preventing 7xxx aluminum alloy hydrogen embrittlement. In this study, the deformation and fracture behaviors of high hydrogen 7xxx alloys containing 0.0% Mn and 0.6% Mn are observed in situ using synchrotron radiation X-ray tomography. The obtained macroscopic hydrogen embrittlement is quantitatively analyzed based on hydrogen partitioning in alloys. Adding 0.6% Mn, generating second-phase particles with high hydrogen trapping abilities, significantly suppresses hydrogen-induced quasicleavage fracture.
Nemoto, Yoshiyuki; Ishijima, Yasuhiro; Kondo, Keietsu; Fujimura, Yuki; Kaji, Yoshiyuki
Journal of Nuclear Materials, 575, p.154209_1 - 154209_19, 2023/03
Previous studies had shown that in certain conditions, the rate of oxidation of zirconium (Zr) based alloy fuel cladding is higher in air-steam mixtures than in dry air. In severe accidents in the spent fuel pool and in other air ingress accidents in nuclear power plants, the cladding is likely to be oxidized in an air-steam mixture, which makes it crucial to have an in-depth understanding of the nature of oxidation and its kinetics in that environment. Oxidation tests were conducted at 800C on Zircaloy-4 specimens in a mix of (air+steam) with various component ratios. Oxidation kinetics, details of the oxide layer, and hydrogen pick-up in the specimen were studied to investigate the mechanism of oxidation in each of these sets of conditions. Zirconium nitride precipitation in the oxide layer during the initial stages of the pre-breakaway oxidation stage and the widespread porous oxide growth on the cladding surface in the latter post-BA oxidation stage are related to the oxidation mechanism in the air-steam mixture. The differences in the mechanism of oxidation of the cladding in dry air and air-steam mixtures are discussed based on the experimental results.
Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi
Materials Transactions, 63(4), p.538 - 544, 2022/04
The time dependence of the corrosion behavior of tantalum (Ta), which is used in nuclear fuel reprocessing equipment, in sodium hydroxide (NaOH) solutions was investigated by immersion tests, and the mechanism of the time dependence was examined via surface observations and electrochemical measurements. The immersion tests were conducted at room temperature with NaOH concentrations ranging from 1 to 7 mol/L for immersion periods of 24 to 168 h. The corrosion rate increased with the NaOH concentration but peaked with the immersion period and then decreased. The time to peak of the corrosion rate was shorter with higher NaOH concentration. The X-ray diffraction (XRD) patterns and Raman spectra of the surfaces of the specimens immersed in the 7 mol/L NaOH solution for more than 48 h showed NaTaO formation. The polarization resistance decreased with immersion time for all NaOH concentrations up to about 24 h after immersion. Thereafter, the polarization resistance increased with immersion time in 7 mol/L NaOH solution and remained almost constant in the other NaOH concentrations. Findings suggested that the change in the corrosion rate was affected by the film formation during immersion, since the time dependence of the polarization resistance and the sum of film resistance and charge transfer resistance had the same tendencies. The precipitation film was mainly NaTaO formed by the dissolution of the passivity film on Ta.
Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi
Zairyo To Kankyo, 70(6), p.192 - 198, 2021/06
The time dependence of corrosion behavior on tantalum used in nuclear fuel reprocessing equipment in sodium hydroxide solution was investigated by immersion corrosion tests, and the mechanism of aging change was discussed from surface observations and electrochemical measurements. The immersion tests were carried out at room temperature with NaOH concentrations ranging from 1 to 7 mol/L and immersion times ranging from 24 to 168 hr, respectively. The corrosion rate increased with NaOH concentration, but peaked with immersion time and then decreased. The time to peak of corrosion rate was shorter with higher NaOH concentration. The SEM observations and Raman analysis at the surface of the specimens that were cleaned and weighed after the immersion test did not show any film formation. On the other hand, the polarization resistance showed a constant value or an increase after a decrease immediately after immersion. It is suggested that the change in corrosion rate is affected by the formation of film by immersion, since the value of polarization resistance is almost the same as the sum of film resistance and charge transfer resistance. The film was considered to be mainly NaTaO formed by the dissolution of Ta.
Hashikura, Yasuaki*; Ishijima, Yasuhiro; Nakahara, Masaumi; Sano, Yuichi; Ueno, Fumiyoshi; Abe, Hitoshi
Hozengaku, 19(3), p.95 - 102, 2020/10
A plutonium concentrator was selected, and constant load tensile tests with controlled applied potentials and electrochemical tests were conducted in nitric acid and sodium nitrate solutions. From the results, a map which shows the effect of nitric acid concentration to crack initiation potential was drawn. And, it was pointed out that not only the nitric acid but also the nitrate ion coordinated to the nitrate must be considered in evaluating the possibility of stress corrosion cracking.
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) 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.
Kato, Chiaki; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Yamamoto, Masahiro
Journal of Nuclear Science and Technology, 53(9), p.1371 - 1379, 2016/09
The effects of crystal textures and the potentials in the anodic oxidation of zirconium in a boiling nitric acid solution were investigated to study the stress corrosion cracking of zirconium in nitric acid solutions. The growth of the zirconium oxide film dramatically changed depending on the applied potential at a closed depassivation potential (1.47 V vs. SSE). At 1.5 V, the zirconium oxide film rapidly grows, and its growth exhibits cyclic oxidation kinetics in accordance with a nearly cubic rate law. The zirconium oxide film grows according to the quantity of electric charge, and the growth rate does not depend on the crystal texture in the pretransition region before the cyclic oxidation kinetics. However, the growth and cracking under the thick oxide film depend on the crystal texture in the transition region. On the normal direction side, the oxide film thickness decreases on average since some areas of the thick oxide film are separated from the specimen surface owing to the cracks in the thick oxide. On the rolling direction side, cracks are found under the thick oxide film, which deeply propagate along the RD without an external stress. The cracks under the thick oxide film propagate to the center of the oxide layer. The cracks in the oxide layer propagate in the (0002)Zr plane in the zirconium matrix. The oxide layer consists of string-like zirconium oxide and zirconium hydride. The string-like zirconium oxide contains orthorhombic ZrO in addition to monoclinic ZrO. As one assumption for the mechanism of crack initiation and propagation without an external stress, it is considered that the oxidizing zirconium hydrides precipitate in the (0002)Zr and then the phase transformation from orthorhombic ZrO to monoclinic ZrO in the oxide layer causes the crack propagation in the (0002) plane.
Ishijima, Yasuhiro; Ueno, Fumiyoshi
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 4 Pages, 2015/05
In this study, to evaluate the effect of thermal aging on creep properties of Alloy 625, we carried out creep tests on aged and solution-treated Alloy 625 at 1073 K. According to the creep test results, time-to-rupture decreased by thermal aging when test stress was more than 100 MPa, but did not change when test stress was less than 100 MPa for any specimens. In the solution-treated alloy, creep deformation behaviors changed over 100 MPa. These results show that time-to-rupture was constant because intermetallic compounds precipitated when the test stress was less than 100 MPa in solution-treated alloy. The observed relationship between creep strain rate and test time showed that the precipitation started after 100 hr for solution treated alloys. These results suggest that intermetallic compounds precipitate immediately after furnace operation. And it is appropriate to use creep data of thermal-aged Alloy 625 for the reducing roasting furnace lifetime prediction.
Ishijima, Yasuhiro; Kato, Chiaki; Motooka, Takafumi; Yamamoto, Masahiro; Kano, Yoichi*; Ebina, Tetsunari*
Materials Transactions, 54(6), p.1001 - 1005, 2013/06
Zirconium has been utilized in nuclear fuel reprocessing plants because of its superior corrosion resistance in nitric acid solutions. However, stress corrosion cracking (SCC) susceptibility of zirconium has been reported in boiling nitric acid solutions at the passivity breakdown potential. However, it has not been clear the SCC initiation and propagation behavior of zirconium. In this study, to clarify the SCC initiation and propagation behavior of zirconium, constant load tensile tests were carried out in boiling nitric acid solutions. From the results, many cracks were initiated under the oxide film and maximum crack led to rupture in the potentials that nobler than passivity breakdown potential. These results showed that the SCC of zirconium in boiling nitric acid solutions is due to the oxide formation. And this SCC behavior suggests that the SCC behavior of zirconium can be attributed to tarnish rupture model.
Komatsu, Atsushi; Ishijima, Yasuhiro; Motooka, Takafumi; Ueno, Fumiyoshi; Yamamoto, Masahiro
Zairyo To Kankyo, 62(5), p.198 - 203, 2013/05
Reduction mechanism of nitrate ion on titanium electrode was investigated using electrochemical method. Cathodic polarization curve of titanium was measured in nitric acid with different concentrations of ions (H,NO,HNO). Reduction mechanism of nitrate ion on titanium was investigated from Tafel slope and reaction order of each ions. It was considered that reduction of titanium oxide was involved in reduction mechanism of nitrate ion on titanium, and reduction mechanism was suggested as follows. NO NO (QE), TiO + H + e TiOOH (QE), NO + TiOOH (NO - TiOOH) (RDS), (NO - TiOOH) + H + e NO + TiO + HO
Kato, Chiaki; Ishijima, Yasuhiro; Yamamoto, Masahiro
Zairyo To Kankyo, 61(1), p.22 - 28, 2012/01
Susceptibility to stress corrosion cracking (SCC) of zirconium was investigated as for in the spent nuclear fuel dissolver environment of the reprocessing facilities. Constant load tensile tests were conducted in the nonradioactive simulated spent nuclear fuel solution in both nobler potential and boiling heat-transfer conditions. It was found that susceptibility to SCC of zirconium strongly depended on electrode potential. The time to failure clearly declined at 1.55V and the cleavage fracture like a facet-shaped was observed. Quasi-cleavage fracture was only observed on the specimen surface less noble than 1.50V, but the time to failure scarcely declined comparing to that of silicon oil. Decline of time to failure was also observed under boiling heat-transfer condition. However, the decline of time to failure under boiling heat-transfer condition was nearly equal to the corresponding temperature to heat-transfer condition with silicon oil. As for the index of susceptibility to SCC, the ratio of transition time from secondary to third creep to failure time indicated that the susceptibility to SCC were very high in 1.55V.
Ioka, Ikuo; Ishijima, Yasuhiro; Usami, Koji; Sakuraba, Naotoshi; Kato, Yoshiaki; Kiuchi, Kiyoshi
Journal of Nuclear Materials, 417(1-3), p.887 - 891, 2011/10
Fe-25Cr-35Ni EHP alloy was developed with conducting the countermeasure for IASCC. It is composed to adjust major elements, to remove harmful impurities and so on. The specimens were irradiated at 553 K for 25000h using JRR-3. The fluence was estimated to be 1.510n/m. Type 304SS was also irradiated as a comparison material. SSRT test was conducted in oxygenated water at 561 K in 7.7 MPa. The fracture mode of EHP alloy was ductile. IGSCC was not observed in the fracture surface. On the other hand, the fraction of IGSCC on the fracture surface of type 304 was about 70%. Microstructural evolution of EHP and type 304 after irradiation was examined by TEM. The defects induced by irradiation mostly consisted of black dots and frank loops in both specimens. No void was also observed in grain and grain boundary of both specimens. There was a little difference in microstructure after irradiation. It is believed that EHP alloy is superior to type 304 in irradiation.
Kato, Chiaki; Ishijima, Yasuhiro; Motooka, Takafumi; Yamamoto, Masahiro
Proceedings of International Conference on Advanced Nuclear Fuel Cycle; Sustainable Options & Industrial Perspectives (Global 2009) (CD-ROM), p.292 - 297, 2009/09
Zirconium has excellent corrosion resistance in nitric acid solutions. However, it has been known that zirconium has stress corrosion cracking (SCC) susceptibility in concentrated HNO with nobler corrosion potential. In this study, we investigated the oxide film growth of zirconium related SCC initiation with various potentials in boiling nitric acid solutions. Electrochemical tests and corrosion tests with various applied potentials conducted in boiling 3, 6 and 9 mol/dm HNO. The potentials in the corrosion tests were set at 1.16, 1.4 and 1.5 V vs. saturated KCl-Ag/AgCl electrode (SSE). These were in the region of trans-passive state of zirconium in boiling nitric acid solution. The test durations were 10, 100 and 500 h. After the corrosion tests, cross-sectional observations of oxide films were conducted. From the results of electrochemical tests of zirconium, passivity region of anodic polarization curves was observed from rest potential to about 1.5 V in boiling 3 mol/dm HNO. Rapid increase of current density was observed at the potential attributed to transition from passivity to trans-passive region. The transition potential in boiling 3, 6 and 9 mol/dm HNO was 1.57, 1.46 and 1.38 V vs. SSE, respectively. The potential was shifted to nobler with decreasing nitric acid concentration. The corrosion tests with various applied potentials indicated that the surfaces of the samples in trans-passive region covered with thick black oxide films. The existence of these oxide films coincided to SCC occurrence. Besides, the oxide film was extremely thin and hardly grew in the passive state potential with no SCC. These results show that SCC of zirconium was initiated by thick oxide film formation. Thus SCC occurrence of zirconium is considerably little in extremely slow oxide film growth rate under the passivity potential.
Ishijima, Yasuhiro; Ioka, Ikuo; Kiuchi, Kiyoshi; Kaneko, Tetsuji*; Okubo, Tsutomu; Yamamoto, Masahiro
Atsuryoku Gijutsu, 47(1), p.12 - 17, 2009/01
We investigate one of these innovative water reactors; Fast Spectrum Light Water Reactor (FLWR). It has unique construction for the reactor core but the fuel cladding material will be exposed in high internal pressure and axial load and complex temperature distribution. Therefore, we conducted a specially designed fatigue-creep test that were simulated several parameters (thermal distribution, temperature variation, internal pressure variation and binding stress) to evaluate an applicability of fuel cladding material for FLWR. Zircalloy-2, which is common cladding material, was used for the test. Test result was confirmed to compare the deformation value between tested and calculated. The result showed the evaluation method could be controlled several parameters simultaneously and the deformation value after the test coincided to the calculated value. This method is sufficient to evaluate thermal deformation characteristics for FLWR.
Kiuchi, Kiyoshi; Ioka, Ikuo; Tanabe, Makoto*; Nanjo, Yoshiyasu*; Ogawa, Hiroaki; Ishijima, Yasuhiro; Tsukatani, Ichiro; Ochiai, Takamasa; Kizaki, Minoru; Kato, Yoshiaki; et al.
JAEA-Research 2006-023, 173 Pages, 2006/03
The research concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100 GWd/t of BWR was pursued for 5 years from 2001 to 2005. On the Phase 1, the modified stainless steel of Fe-25Cr-35Ni-0.2Ti as fuel claddings and Nb-Mo alloy as a liner for inhibiting the pellet- clad interaction were selected as candidate materials, by evaluating fundamental properties required to BWR cladding materials, that are the nuclear economy, radioactivity, mass-transfer, irradiation properties, mechanical properties so on. On the present study, the making process of cladding tubes, lining by diffusion bonding, end plug by laser welding were developed and optimized, by considering the practical use of fuel elements consists of these candidates. The practical applicability was basically examined by irradiation tests using the accelerator of TIARA and the research reactor of JRR-3, for mainly confirming the resistance to IGSCC as one of the current important issues of BWR core materials of low carbon grade stainless steels. Creep and fatigue testing data were also obtained for evaluating the long performance of candidate materials. The behavior as fuel elements was analyzed with the safety calculation code for BWRs. The obtained results were established as a data base system, by considering the applicability to the fuel design and in-pile loop tests.
Arai, K.*; Ninomiya, Akira*; Ishigooka, Takeshi*; Takano, Katsutoshi*; Nakajima, Hideo; Michael, P.*; Vieira, R.*; Martovetsky, N.*; Sborchia, C.*; Alekseev, A.*; et al.
Cryogenics, 44(1), p.15 - 27, 2004/01
no abstracts in English
Koizumi, Norikiyo; Azuma, Katsunori*; Tsuchiya, Yoshinori; Matsui, Kunihiro; Takahashi, Yoshikazu; Nakajima, Hideo; Nishijima, Gen; Nunoya, Yoshihiko; Ando, Toshinari; Isono, Takaaki; et al.
Fusion Engineering and Design, 58-59, p.1 - 5, 2001/11
no abstracts in English
Kato, Takashi; Tsuji, Hiroshi; Ando, Toshinari; Takahashi, Yoshikazu; Nakajima, Hideo; Sugimoto, Makoto; Isono, Takaaki; Koizumi, Norikiyo; Kawano, Katsumi; Oshikiri, Masayuki*; et al.
Fusion Engineering and Design, 56-57, p.59 - 70, 2001/10
no abstracts in English
Tsuji, Hiroshi; Okuno, Kiyoshi*; Thome, R.*; Salpietro, E.*; Egorov, S. A.*; Martovetsky, N.*; Ricci, M.*; Zanino, R.*; Zahn, G.*; Martinez, A.*; et al.
Nuclear Fusion, 41(5), p.645 - 651, 2001/05
no abstracts in English
Takahashi, Yoshikazu; Ando, Toshinari; Hiyama, Tadao; Nakajima, Hideo; Kato, Takashi; Sugimoto, Makoto; Isono, Takaaki; Oshikiri, Masayuki*; Kawano, Katsumi; Koizumi, Norikiyo; et al.
Teion Kogaku, 35(7), p.357 - 362, 2000/07
no abstracts in English