Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi
Nippon 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
Takahashi, Yoshikazu; Nunoya, Yoshihiko; Nishijima, Gen; Koizumi, Norikiyo; Matsui, Kunihiro; Ando, Toshinari; Hiyama, Tadao; Nakajima, Hideo; Kato, Takashi; Isono, Takaaki; et al.
IEEE Transactions on Applied Superconductivity, 10(1), p.580 - 583, 2000/03
no abstracts in English
Ioka, Ikuo; Ishijima, Yasuhiro; Kiuchi, Kiyoshi; Kizaki, Minoru; Kato, Yoshiaki; Fujimura, Ken*; Obata, Hiroyuki*
no journal, ,
The IASCC and loss of ductility concerning new cladding materials for ultra-high burnup of fuel elements with MOX fuels aiming at 100GWd/t of BWR was performed after neutron irradiation. The specimens machined from tubes made of Fe-25Cr-35Ni-0.2Ti UHP and SUS304 were irradiated at 290C up to 1.8dpa(1.510n/m) in JRR-3. The ultimate tensile strength of both specimens increased and the fracture elongation decreased in tensile test at 288C. The loss of ductility was almost equal with data of previous literatures. IASCC was recognized on SUS304 by SSRT testing under high temperature water, but not on Fe-25Cr-35Ni-0.2Ti UHP.
Ishijima, Yasuhiro; Ioka, Ikuo; Kiuchi, Kiyoshi; Usami, Koji; Kato, Yoshiaki; Fujimura, Ken*
no journal, ,
no abstracts in English
Motooka, Takafumi; Yamamoto, Masahiro; Ueno, Fumiyoshi; Kato, Chiaki; Ishijima, Yasuhiro
no journal, ,
It is well known that stainless steels rapidly corrode in nitric acid solutions included oxidizing ions. But the time effect of oxidizing ions on corrosion has been rarely studied. In this repot, the change in corrosion rate with time was investigated using Cr(VI) and V(V) species. Corrosion test of SUS304L stainless steel was conducted in boiling nitric acid solutions. Quantitative analysis of test solutions was done before and after corrosion tests. Corrosion rate of stainless steel in high concentration nitric acid solution with Cr(VI) was rapidly decreased and Cr(VI) was reduced to Cr(III). Corrosion rate of stainless steel in nitric acid solution with V(V) and the amount of V(V) were not greatly changed with time. It was concluded that the corrosion of a stainless steel in nitric acid solutions included Cr(VI) was greatly affected by concentration of nitric acid and test time, and the corrosion of a stainless steel in nitric acid solutions included V(V) was not affected by test time.
Yamamoto, Masahiro; Kato, Chiaki; Ishijima, Yasuhiro; Kano, Yoichi*; Ebina, Tetsunari*
no journal, ,
Zirconium had been used in nuclear fuel reprocessing plants, because of its superior corrosion resistance in nitric acid solution. However, stress corrosion cracking (SCC) was reported at highly concentrated nitric acid solution. In this report, electrochemical measurements were utilized to clarify the criteria of SCC occurrence of zirconium in boiling nitric acid solutions. For the relations between potentials and SCC susceptibilities, SCC behavior of zirconium was attributed to tarnish rupture model. The observations of fractured surfaces and cross sections of cracks also supported the validity of the model. SCC susceptibility of zirconium was mainly affected by the thick oxide film formation at the surface. In our whole observations, SCC of zirconium occurred only at transpassive regions. When a reprocessing plant made of zirconium operates in the passivity state, there is no SCC occurrence in the equipment.