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Journal Articles

Effect of $$B$$-site randomness on the antiferroelectric/relaxor nature of the ground state; Diffuse and inelastic X-ray scattering study of Pb(In$$_{1/2}$$Nb$$_{1/2}$$)O$$_3$$

Owada, Kenji*; Tsukada, Shinya*; Fukuda, Tatsuo; Tsutsui, Satoshi*; Baron, A. Q. R.*; Mizuki, Junichiro*; Owa, Hidehiro*; Yasuda, Naohiko*; Terauchi, Hikaru*

Physical Review B, 98(5), p.054106_1 - 054106_10, 2018/08

https://doi.org/10.1103/PhysRevB.98.054106
 Times Cited Count：3 Percentile：15.56(Materials Science, Multidisciplinary)

Journal Articles

Negative correlation between electrical response and domain size in a Ti-composition-gradient Pb[(Mg$$_{1/3}$$Nb$$_{2/3}$$)$$_{1-x}$$Ti$$_{x}$$]O$$_{3}$$ crystal near the morphotropic phase boundary

Shimizu, Daisuke*; Tsukada, Shinya*; Matsuura, Masato*; Sakamoto, Junya*; Kojima, Seiji*; Namikawa, Kazumichi*; Mizuki, Junichiro; Owada, Kenji

Physical Review B, 92(17), p.174121_1 - 174121_5, 2015/11

https://doi.org/10.1103/PhysRevB.92.174121
 Times Cited Count：13 Percentile：50.04(Materials Science, Multidisciplinary)

The phase diagram and the relationship between the crystal coherence length and electrical response of Pb[(Mg$$_{1/3}$$Nb$$_{2/3}$$)$$_{1-x}$$Ti$$_{x}$$]O$$_{3}$$ (PMN-xPT) near the morphotropic phase boundary (MPB) have been precisely investigated using a single crystal with a Ti composition gradient by synchrotron X-ray diffraction and inelastic light scattering at room temperature. The crystal has two boundaries at Ti compositions of 29.0 mol% and 34.7 mol% which correspond to the phase boundaries between the monoclinic B (MB) and C (MC) phases and between the MC and tetragonal (T) phases, respectively. It is shown that there is a strong negative correlation between the electrical response and the crystal coherence length at the sub-$$mu$$m scale. The results are explained by the size effects of domains near the MPB.

Journal Articles

Effect of $$gamma$$-ray irradiation on corrosion of low alloy steel in neutral water

Yamamoto, Masahiro; Nakano, Junichi; Komatsu, Atsushi; Sato, Tomonori; Tsukada, Takashi

Proceedings of 19th International Corrosion Congress (19th ICC) (CD-ROM), 6 Pages, 2014/11

Corrosion protection of RPV and PCV is an important issue for the long term maintenance until the end of the decommissioning procedures. One of the uncertain factors for the issue is an effect of radioactivity on corrosion of LAS and CS. Corrosion tests using LAS and CS were conducted in $$gamma$$-rays irradiated condition. Oxygen and hydrogen peroxide concentrations in the water were measured after the tests. Corrosion test results indicated that the amounts of corrosion increased by $$gamma$$-rays irradiation both air and nitrogen atmosphere. And also corrosion amounts increased with $$gamma$$-ray dose rates. Electrochemical analyses indicated that cathodic reaction of Hydrogen peroxide was controlled by diffusion process. The measured diffusion constant of H$$_{2}$$O$$_{2}$$ was about 0.75 times to that of oxygen. From these results, it is estimated that corrosion of LAS and CS in $$gamma$$-ray irradiated condition was evaluated by the cathodic reduction reaction of oxidant.

Journal Articles

Effects of hydrazine addition and N$$_{2}$$ atmosphere on the corrosion of reactor vessel steels in diluted seawater under $$gamma$$-rays irradiation

Nakano, Junichi; Kaji, Yoshiyuki; Yamamoto, Masahiro; Tsukada, Takashi

Journal of Nuclear Science and Technology, 51(7-8), p.977 - 986, 2014/07

https://doi.org/10.1080/00223131.2014.890917
 Times Cited Count：4 Percentile：30.92(Nuclear Science & Technology)

Seawater was injected into the reactor cores in the Fukushima Daiichi Nuclear Power Station. Corrosion of reactor vessel steels is considered to progress. To evaluate durability of the reactor vessel steels, corrosion tests were conducted in diluted seawater at 50 $$^{circ}$$C under $$gamma$$-rays irradiation. 10 mg/L and 100 mg/L N$$_{2}$$H$$_{4}$$ were added to diluted seawater. Without addition of N$$_{2}$$H$$_{4}$$, weight loss in the vessel steels irradiated with the 0.2 kGy/h dose rate was comparable with those without irradiation and weight loss in the vessel steels irradiated with the 4.4 kGy/h dose rate was higher than those without irradiation. Under irradiation, weight loss in the vessel steels in diluted seawater containing N$$_{2}$$H$$_{4}$$ was comparable with that in diluted seawater without N$$_{2}$$H$$_{4}$$. When gas phase in the flask was replaced with N$$_{2}$$, weight loss in the vessel steels, and O$$_{2}$$ and H$$_{2}$$O$$_{2}$$ concentrations in the diluted seawater decreased.

Journal Articles

Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under $$gamma$$-rays irradiation

Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(1), p.1 - 6, 2014/03

https://doi.org/10.3327/taesj.J13.015

The seawater was injected into reactor cores of Units 1, 2 and 3 in the Fukushima Daiichi Nuclear Power Plant. To investigate effects of $$gamma$$ rays irradiation on corrosion of carbon steel and low alloy steel, corrosion tests were performed in the diluted seawater at 50 $$^{circ}$$C under $$gamma$$ rays irradiation with dose rates of 4.4 kGy/h and 0.2 kGy/h. Hydrazine (N$$_{2}$$H$$_{4}$$) was added to the diluted seawater. In the diluted seawater without N$$_{2}$$H$$_{4}$$, the weight losses of the steels irradiated with 0.2 kGy/h were similar to those of the unirradiated steels, and the weight losses of the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those of the unirradiated steels. The weight losses of the steels irradiated in the diluted seawater containing N$$_{2}$$H$$_{4}$$ were similar to those in the diluted seawater without N$$_{2}$$H$$_{4}$$. When N$$_{2}$$ was introduced to gas phase in the flasks during $$gamma$$ rays irradiation, the weight losses of the steels decreased.

Journal Articles

Effects of temperature on stress corrosion cracking behavior of stainless steel and outer oxide distribution in cracks due to exposure to high-temperature water containing hydrogen peroxide

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Yamamoto, Masahiro; Tsukada, Takashi; Kaji, Yoshiyuki

Journal of Nuclear Materials, 444(1-3), p.454 - 461, 2014/01

https://doi.org/10.1016/j.jnucmat.2013.10.031
 Times Cited Count：3 Percentile：23.92(Materials Science, Multidisciplinary)

Cracking growth tests were conducted in high-temperature water containing hydrogen peroxide (H$$_{2}$$O$$_{2}$$) at 561 to 423 K to evaluate the effects of H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC) of stainless steel (SS) at temperature lower than the boiling water reactor (BWR) operating temperature. Small compact tension (CT) specimens were prepared from thermally sensitized type 304 SS. Despite the observation of only a small portion intergranular SCC (IGSCC) near the side groove of the CT specimen at 561 K in high-temperature water containing 100 ppb of H$$_{2}$$O$$_{2}$$, the IGSCC area expanded to the central region of the CT specimens at 423 and 453 K. Effects of H$$_{2}$$O$$_{2}$$ on SCC appeared intensely at temperature lower than the BWR operating temperature. To estimate the environment in the cracks, outer oxide distribution on the fracture surface and fatigue pre-crack were examined by laser Raman spectroscopy and thermal equilibrium calculation was performed.

Journal Articles

Morphology of stress corrosion cracking due to exposure to high-temperature water containing hydrogen peroxide in stainless steel specimens with different crevice lengths

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Yamamoto, Masahiro; Tsukada, Takashi; Kaji, Yoshiyuki

Journal of Nuclear Materials, 441(1-3), p.348 - 356, 2013/10

https://doi.org/10.1016/j.jnucmat.2013.05.073
 Times Cited Count：2 Percentile：18.63(Materials Science, Multidisciplinary)

Crack growth tests were performed in high-temperature water containing hydrogen peroxide (H$$_{2}$$O$$_{2}$$) to evaluate the relationships between the crevice structure and H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC) growth morphology of stainless steel (SS). Small compact tension (CT) specimens with different fatigue pre-crack lengths were prepared. 20$$sim$$300 ppb H$$_{2}$$O$$_{2}$$ was injected into the high-temperature water at 561 K. Intergranular SCC (IGSCC) was observed only near the side grooves of the CT specimens. Owing to pre-crack shortening, the IGSCC area expanded to the central region of the CT specimens. The effects of H$$_{2}$$O$$_{2}$$ on SS appeared intensely near the surfaces exposed to high levels of H$$_{2}$$O$$_{2}$$. The calculations for the percentage of H$$_{2}$$O$$_{2}$$ remaining showed that the effects of H$$_{2}$$O$$_{2}$$ flowed from both sides of the crack were more obvious than those flowed from the crack mouth.

Journal Articles

Influence of microstructure on IASCC growth behavior of neutron irradiated type 304 austenitic stainless steels in simulated BWR condition

Kaji, Yoshiyuki; Miwa, Yukio*; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

International Journal of Nuclear Energy Science and Engineering, 2(3), p.65 - 71, 2012/09

Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The specimens were irradiated in the core region of the Japan Materials Testing Reactor (JMTR) in simulated BWR water environments at 288 $$^{circ}$$C from 0.37 to 5.55$$times$$10$$^{25}$$ n/m$$^{2}$$ (E$$>$$ 1 MeV) (0.62 to 9.2 dpa). The CGRs of base metals in high electrochemical corrosion potential (ECP) condition with 10 $$<$$ stress intensity factor, K $$<$$ 30 MPam$$^{1/2}$$, increased with increasing neutron fluence until 2 dpa and the CGRs were almost the same from 2 to 10 dpa. We investigated the influence of microstructure on CGR by microstructure observation and local strain measurement around the precipitate. This paper will discuss the relationship between CGR and microstructure, radiation hardening, radiation induced segregation.

Journal Articles

Effects of temperature on SCC propagation in high temperature water injected with hydrogen peroxide

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Kaji, Yoshiyuki; Yamamoto, Masahiro; Tsukada, Takashi

Proceedings of 2012 Nuclear Plant Chemistry Conference (NPC 2012) (CD-ROM), 9 Pages, 2012/09

It has been reported that the corrosion behavior of stainless steels in high temperature water with hydrogen peroxide (H$$_{2}$$O$$_{2}$$) was different from those with O$$_{2}$$. To evaluate the effect of H$$_{2}$$O$$_{2}$$ on stress corrosion cracking (SCC), SCC growth tests were conducted in high temperature water injected with H$$_{2}$$O$$_{2}$$. In 100 ppb H$$_{2}$$O$$_{2}$$ at 561 K, an intergranular SCC (IGSCC) was observed only small portion of area near the side grooves of the CT specimen. In 100 ppb H$$_{2}$$O$$_{2}$$ at 453 K, however, IGSCC extended to the central region of the CT specimen. Effects of H$$_{2}$$O$$_{2}$$ on SCC growth behavior appeared stronger at lower temperature due to a reduction of the thermal decomposition of H$$_{2}$$O$$_{2}$$. Moreover, outer oxide layer of oxide film formed on the crack of the CT specimen was examined to estimate environmental situations in the cracks and a thermal equilibrium calculation was performed.

JAEA Reports

Evaluating techniques and phenomena of Stress Corrosion Cracking (SCC) in Light Water Reactors (LWRs); SCC evaluating techniques for predicting core internal and pipe aging of LWRs, technical data collection (Contract research)

Yamamoto, Masahiro; Kato, Chiaki; Sato, Tomonori; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Kaji, Yoshiyuki; Tsujikawa, Shigeo*; Hattori, Shigeo*; Yoshii, Tsuguyasu*; et al.

JAEA-Review 2012-007, 404 Pages, 2012/03

JAEA-Review-2012-007.pdf:36.72MB

There are many LWRs which have been operated for more than 20 years in Japan and it is expected that technique corresponding to aging plants are necessary established for safety operation in LWRs. A lot of troubles related to SCC are reported and many investigations are concerned with SCC mechanism and technical evaluation. In this paper, those research data were collected as possible widely and reviewed systematically. Current circumstances concerned with SCC in LWRs were reviewed specifically as follows: SCC incidents, SCC evaluation methods for crack initiation and propagation, the investigations concerned with SCC mechanism and monitoring technique for corrosive environment. Influences with reactor types (BWR, PWR), materials (stainless steels, Ni alloys) and SCC evaluating methods (laboratories and actual plants) were summarized as graphs and tables easy to understand in common/difference points concerned with SCC. From these arranged results, future themes were considered and remarked SCC phenomenon was summarized in actual plants. As for SCC evaluations under the accelerate conditions in the laboratory test, it was suggested that a computational prediction and modeling including statistical technique and microscopic analysis in crack initiation were important. Furthermore it was suggested that monitoring techniques predicting SCC initiation and grasping plant circumstance in operation and feasibility in actual plants were important.

Journal Articles

Development of remote welding techniques for in-pile IASCC capsules and evaluation of material integrity on capsules for long irradiation period

Shibata, Akira; Nakano, Junichi; Omi, Masao; Kawamata, Kazuo; Nakagawa, Tetsuya; Tsukada, Takashi

Journal of Nuclear Materials, 422(1-3), p.14 - 19, 2012/03

https://doi.org/10.1016/j.jnucmat.2011.11.057
 Times Cited Count：0 Percentile：0.01(Materials Science, Multidisciplinary)

To simulate Irradiation assisted stress corrosion cracking (IASCC) behavior by in-pile experiments, it is necessary to irradiate specimens up to a neutron fluence that is higher than the IASCC threshold fluence. Pre-irradiated specimens must be relocated from pre-irradiation capsules to in-pile capsules. Hence, a remote welding machine has been developed. And the integrity of capsule housing for a long term irradiation was evaluated by tensile tests in air and slow strain rate tests in water. Two type specimens were prepared. Specimens were obtained from the outer tubes of capsule irradiated to 1.0-3.9 $$times$$ 10$$^{26}$$ n/m$$^{2}$$ (E$$>$$ 1 MeV). And specimens were irradiated in a leaky capsule to 0.03-1.0 $$times$$ 10$$^{26}$$ n/m$$^{2}$$. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.

Journal Articles

SCC susceptibility of cold-worked stainless steel with minor element additions

Nakano, Junichi; Nemoto, Yoshiyuki; Tsukada, Takashi; Uchimoto, Tetsuya*

Journal of Nuclear Materials, 417(1-3), p.883 - 886, 2011/10

https://doi.org/10.1016/j.jnucmat.2010.12.151
 Times Cited Count：2 Percentile：18.29(Materials Science, Multidisciplinary)

To examine the effects of minor elements on stress corrosion cracking (SCC) susceptibility of low carbon stainless steels with work hardened layer, a high purity type 304 stainless steel was fabricated and minor elements, Si, S, P, C or Ti, were added. Work hardened layer was introduced by shaving on the surface of stainless steels. The specimens were exposed to a boiling 42% MgCl$$_{2}$$ solution for 20 hours and the number and the length of initiated cracks were examined. SCC susceptibility of the specimen with P was the highest and that of the specimen with C was the lowest in all specimens. By magnetic force microscope examination, magnetic phase expected to be martensitic phase was detected near surface. Since corrosion resistance of martensite is lower than that of austenite, the minor elements additions would affect SCC susceptibility through the amount of the transformed martensite, i.e. austenite stability.

Journal Articles

Effects of hydrogen peroxide on SCC behavior of type 304 stainless steel under simulated water radiolysis environment

Nakano, Junichi; Sato, Tomonori; Kato, Chiaki; Yamamoto, Masahiro; Tsukada, Takashi

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10

To evaluate the effects of hydrogen peroxide (H$$_{2}$$O$$_{2}$$) on stress corrosion cracking (SCC) behavior of stainless steels (SSs), crack growth tests and in-situ electrochemical impedance spectroscopy (EIS) in high temperature water injected with H$$_{2}$$O$$_{2}$$ were carried out. On the fracture surface of a compact tension (CT) specimen exposed 100 ppb H$$_{2}$$O$$_{2}$$ at 561 K, intergranular SCC (IGSCC) was observed only near side grooves of the CT specimen. To increase the amount of remaining H$$_{2}$$O$$_{2}$$, Chevron notches were eliminated from a CT specimen. Consequently expansion of IGSCC area was observed into the central region of the CT specimen. EIS was conducted on SS creviced corrosion specimens. Using the obtained charge transfer resistance, ratio of remaining H$$_{2}$$O$$_{2}$$ inside crevice was calculated with a differential equation code. Ratio of remaining H$$_{2}$$O$$_{2}$$ in the CT specimen without Chevron notches was calculated as higher than that of the one with Chevron notches.

Journal Articles

Stress corrosion cracking behavior of type 304 stainless steel irradiated under different neutron dose rates at JMTR

Kaji, Yoshiyuki; Kondo, Keietsu; Aoyagi, Yoshiteru; Kato, Yoshiaki; Taguchi, Taketoshi; Takada, Fumiki; Nakano, Junichi; Ugachi, Hirokazu; Tsukada, Takashi; Takakura, Kenichi*; et al.

Proceedings of 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (CD-ROM), p.1203 - 1216, 2011/08

In order to investigate the effect of neutron dose rate on tensile property and irradiation assisted stress corrosion cracking (IASCC) growth behavior, the crack growth rate (CGR) test, tensile test and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating the temperature of boiling water reactor (BWR) up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). The radiation hardening increased with the dose rate, but there was little effect on CGR. Increase of the yield strength of specimens irradiated with the low dose rate condition was caused by the increase of number density of frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, there was little effect on local plastic deformation behavior near crack tip in the crystal plasticity simulation.

Journal Articles

The Electrochemical impedance spectroscopy of stainless steel exposed to high temperature water containing hydrogen peroxide

Sato, Tomonori; Noda, Kazuhiko*; Kato, Chiaki; Yamamoto, Masahiro; Nakano, Junichi; Tsukada, Takashi

Proceedings of Symposium on Water Chemistry and Corrosion in Nuclear Power Plants in Asia 2009 (CD-ROM), p.232 - 237, 2009/10

In this work, to clarify the electrochemical behaviors at the surfaces of stainless steels (SSs) in high temperature water containing hydrogen peroxide (H$$_{2}$$O$$_{2}$$), the in-situ electrochemical impedance spectroscopy (EIS) of SSs exposed to high temperature water was carried out. The materials of test specimens were type 316L SS and type 304L SS. The range of the applied frequency in EIS was 100 kHz to 1 m or 10 mHz. The charge transfer resistance at the boundary between the oxide film and the base metal (R$$_{rm ct}$$) in oxygen (O$$_{2}$$) condition was larger than R$$_{rm ct}$$ in H$$_{2}$$O$$_{2}$$ condition. This indicates that the corrosion rate of type 316L SS in high temperature water containing H$$_{2}$$O$$_{2}$$ is larger than that in O$$_{2}$$ contained water. The R$$_{rm ct}$$ of type 316L SS was larger than that of type 304L SS in high temperature water containing H$$_{2}$$O$$_{2}$$. This indicates that the corrosion resistance of type 316L SS is higher than that of type 304L SS.

Journal Articles

Influence of microstructure on IASCC growth behavior of neutron irradiated type 304 austenitic stainless steels in simulated BWR condition

Kaji, Yoshiyuki; Miwa, Yukio; Shibata, Akira; Nakano, Junichi; Tsukada, Takashi; Takakura, Kenichi*; Nakata, Kiyotomo*

Proceedings of 14th International Conference on Environmental degradation of Materials in Nuclear Power Systems (CD-ROM), p.1181 - 1191, 2009/08

The CGR tests of neutron irradiated Type 304 SS were conducted in BWR conditions and the results were compared with those of Type 304L and 316L SS, and following results were obtained. (1) The CGR increase with increasing neutron fluence and the power law of K on the CGR was observed above F2 neutron fluence level (1.4 dpa). The different tendency is observed between Type 304 SS and L-grade SS (Type 304L and 316L SS) with increasing neutron fluence above F3 (4.3 dpa) level. (2) The CGR of Type 304 SS is slightly small as compared with those of Type 304L and 316L SS at the same neutron fluence and shows an increasing tendency above 4 dpa and reaches to 1.0$$times$$10$$^{-9}$$m/s in 9 dpa. (3) The neutron fluence dependence on uniform elongation is different with Type 304, 304L SS and Type 316L SS, that is, the neutron fluence in which the local deformation like channeling deformation is dominant, is high for Type 316L SS.

Journal Articles

Examination of relation between IASCC susceptibility and magnetic property

Takaya, Shigeru; Nagae, Yuji; Yoshitake, Tsunemitsu; Nemoto, Yoshiyuki; Nakano, Junichi; Ueno, Fumiyoshi; Aoto, Kazumi; Tsukada, Takashi

E-Journal of Advanced Maintenance (Internet), 1(1), p.44 - 51, 2009/05

As the result of comparing the magnetic flux density and the IASCC susceptibility evaluated by SSRT test on neutron irradiated model alloys, it was shown that there is the relation without depending on dose level and chemical compositions as long as the contribution of neutron irradiation to SCC was seen. Furthermore, measuring the magnetic flux density of unirradiated simulated degraded materials indicates that not only change in chemical compositions but also some defects are needed for the magnetic flux density to increase. These results show the possibility of non-destructive estimation of susceptibility to IASCC by measuring magnetic flux density.

JAEA Reports

Development on crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation, 1 (Contract research)

Izumo, Hironobu; Chimi, Yasuhiro; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; et al.

JAEA-Technology 2009-011, 31 Pages, 2009/04

JAEA-Technology-2009-011.pdf:4.38MB

Regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) for austenitic stainless steel of the light water reactor (LWR), a lot of data that concerns the post irradiation evaluation (PIE) is acquired. However, IASCC occurs in LWR condition. Therefore, it is necessary to confirm adequacy of the PIE data comparing the experiment data under the simulated LWR condition. Bigger specimen is needed to acquire the effective data for the destruction dynamics in the study of stress corrosion cracking under neutron irradiation condition. Therefore, development of a new crack growth unit which can load to bigger is necessary to the neutron irradiation test. As a result, a prospect was provided in the unit that could load to specimen by changing load mechanism to the lever type from the linear type. And also, in the development of crack propagation unit, some technical issues were extracted from the discussion of the unit structure adopting the LVDT (Linear Variable Differential Transformer).

Journal Articles

In-situ SCC observation of thermally-sensitized and cold-worked type 304 stainless steel irradiated to a neutron fluence of 1$$times$$10$$^{25}$$n/m$$^{2}$$

Nakano, Junichi; Nemoto, Yoshiyuki; Miwa, Yukio; Usami, Koji; Tsukada, Takashi; Hide, Koichiro*

Journal of Nuclear Materials, 386-388, p.281 - 285, 2009/04

https://doi.org/10.1016/j.jnucmat.2008.12.114
 Times Cited Count：4 Percentile：30.49(Materials Science, Multidisciplinary)

Crack initiation and crack growth processes of irradiation assisted stress corrosion cracking on stainless steels were studied by slow strain rate testing in oxygenated high temperature water at 561 K. In-situ observation was carried out during SSRT. Specimens of type 304 stainless steel were subjected to a solution annealing (SA), a thermally sensitization (TS), or a cold working (CW) and irradiated to 1.0$$times$$10$$^{25}$$ n/m$$^{2}$$ (E $$>$$ 1 MeV) at 323 K in the Japan Material Testing Reactor (JMTR). Crack initiations were observed before the maximum stress would be reached for the CW material in in-situ observation. In fracture surface examination, the TS material exhibited almost intergranular stress corrosion cracking while mixtures of transgranular stress corrosion cracking and ductile dimple fracture were observed for the SA and the CW materials.

Journal Articles

Observation of high-$$j$$ quasiparticle states in $$^{249}$$Cm by in-beam $$gamma$$-ray spectroscopy using heavy-ion transfer reactions

Ishii, Tetsuro; Makii, Hiroyuki*; Asai, Masato; Tsukada, Kazuaki; Toyoshima, Atsushi; Matsuda, Makoto; Makishima, Akiyasu*; Shigematsu, Soichiro*; Kaneko, Junichi*; Shizuma, Toshiyuki; et al.

Physical Review C, 78(5), p.054309_1 - 054309_11, 2008/11

https://doi.org/10.1103/PhysRevC.78.054309
 Times Cited Count：14 Percentile：64.91(Physics, Nuclear)

83 (Records 1-20 displayed on this page)