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Watanabe, So; Takahatake, Yoko; Ogi, Hiromichi*; Osugi, Takeshi; Taniguchi, Takumi; Sato, Junya; Arai, Tsuyoshi*; Kajinami, Akihiko*
Journal of Nuclear Materials, 585, p.154610_1 - 154610_6, 2023/11
Times Cited Count:0 Percentile:0.02Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Murakami, Tatsutoshi
Journal of Nuclear Materials, 584, p.154576_1 - 154576_11, 2023/10
Times Cited Count:0 Percentile:0.02(Materials Science, Multidisciplinary)Ukai, Shigeharu; Sakamoto, Kan*; Otsuka, Satoshi; Yamashita, Shinichiro; Kimura, Akihiko*
Journal of Nuclear Materials, 583, p.154508_1 - 154508_24, 2023/09
Times Cited Count:0 Percentile:0.74(Materials Science, Multidisciplinary)Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki
Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08
Times Cited Count:0 Percentile:0.02(Materials Science, Multidisciplinary)Vauchy, R.; Sunaoshi, Takeo*; Hirooka, Shun; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07
Times Cited Count:1 Percentile:86.79(Materials Science, Multidisciplinary)Kitagaki, Toru; Krasnov, V.*; Ikeda, Atsushi
Journal of Nuclear Materials, 576, p.154224_1 - 154224_14, 2023/04
Times Cited Count:0 Percentile:0.74(Materials Science, Multidisciplinary)Nemoto, Yoshiyuki; Ishijima, Yasuhiro; Kondo, Keietsu; Fujimura, Yuki; Kaji, Yoshiyuki
Journal of Nuclear Materials, 575, p.154209_1 - 154209_19, 2023/03
Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)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.
Miyazawa, Takeshi; Kikuchi, Yuta*; Ando, Masami*; Yu, J.-H.*; Yabuuchi, Kiyohiro*; Nozawa, Takashi*; Tanigawa, Hiroyasu*; Nogami, Shuhei*; Hasegawa, Akira*
Journal of Nuclear Materials, 575, p.154239_1 - 154239_11, 2023/03
Times Cited Count:0 Percentile:0.02(Materials Science, Multidisciplinary)Yabuuchi, Kiyohiro*; Suzudo, Tomoaki
Journal of Nuclear Materials, 574, p.154161_1 - 154161_6, 2023/02
Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)In nuclear materials, irradiation defects cause degradation of mechanical properties. In these materials, the relationship between dislocations and voids is particularly important for mechanical strength. Although only spherical voids have been studied in the past, this study focuses on faceted voids, which are observed simultaneously with spherical voids. In the current study, molecular dynamics was used to analyze the effect of faceted voids in the irradiation hardening of pure iron. Specifically, we clarified the difference in obstacle strength and interaction processes between spherical voids and faceted voids, and that even faceted voids show differences in interaction depending on their crystallographic arrangement with dislocations.
Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01
Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)Oka, Hiroshi*; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Hashimoto, Naoyuki*
Journal of Nuclear Materials, 572, p.154032_1 - 154032_8, 2022/12
Times Cited Count:2 Percentile:68.58(Materials Science, Multidisciplinary)9Cr oxide dispersion strengthened steels with slightly different nitrogen concentrations (0.0034 - 0.029 wt%) were prepared and their creep property at 973 K was investigated with microstructural characterization before and after the creep test. The creep strength decreased significantly as the nitrogen concentration increased. Microstructural observation revealed that, in the higher nitrogen concentration specimen, coarse Y-rich inclusions were found along the boundary between transformed ferrite region and residual ferrite region. The solubility difference of nitrogen in and
phase would induce the localized increment of nitrogen concentration in the boundary region during the austenitizing process, resulting in the thermodynamic destabilization and subsequent coarsening of the dispersed oxide particles. The rows of creep voids were found near the rupture part of the crept specimen, suggesting that the coarse inclusions were the starting point of creep void formation and the subsequent premature fracture.
Galvin, C. O. T.*; Machida, Masahiko; Nakamura, Hiroki; Andersson, D. A.*; Cooper, M. W. D.*
Journal of Nuclear Materials, 572, p.154028_1 - 154028_8, 2022/12
Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)UO is the primary conventional fuel used in most nuclear reactors with Gd
O
commonly added as a burnable absorber to produce a more level power distribution in the reactor core at the beginning of operation. It can also be mixed with other actinide oxides to produce mixed oxide (MOx) fuel. In this study, molecular dynamics simulations were used to predict the specific heat capacity of Gd-doped PuO
, UO
and (U,Pu)O
MOx accommodating Gd
substituted at cation sites via two charge compensation mechanisms - oxygen vacancy formation and the oxidation of U
to U
. The specific heat capacity values for PuO
and UO
are in good agreement with other studies showing a distinct peak at high temperatures - above 1800 K. As Gd
is added, the peak height reduces for each composition considered. An analytical fit was applied to the data where Gd
was fully charge compensated by either oxygen vacancies or U
. The expression was then validated by predicting the specific heat capacity for three compositions of (U
Pu
)
Gd
O
containing both oxygen vacancies and U
, and compared to molecular dynamics data.
Okamoto, Yoshihiro; Shiwaku, Hideaki; Shimamura, Keisuke*; Kobayashi, Hidekazu; Nagai, Takayuki; Inose, Takehiko*; Sato, Seiichi*; Hatakeyama, Kiyoshi*
Journal of Nuclear Materials, 570, p.153962_1 - 153962_13, 2022/11
Simulated nuclear waste glass samples containing phosphorus, which increase the solubility of molybdenum, were prepared and analyzed using synchrotron X-ray Absorption Fine Structure (XAFS) analysis for some constituent elements and Raman spectroscopic analysis of their complex structure. Changes in local structure and chemical state due to different phosphorus additions and waste loading rates were systematically studied. Consequently, no crystalline phase due to the molybdate compound was observed even at a maximum waste content of 30 wt% (corresponding to 1.87 mol% MoO). Oxidation proceeded when the waste-loading rate was increased, whereas the reduction proceeded when phosphorus was added. In some cases, the effects of oxidation and reduction were offset. The local structure around specific elements can be classified as follows; Zn that is affected mainly by the waste-loading rate, Ce that is affected by both the waste-loading rate and phosphorus addition, and Zr element that is not affected by either of them. From the comparison between the analytical results of Mo and other elements, it was considered that the added phosphorus exists as a free PO
structural unit and may deprive the alkali metal coordinated to the molybdate ion.
Ito, Daisuke*; Sato, Hirotaka*; Odaira, Naoya*; Saito, Yasushi*; Parker, J. D.*; Shinohara, Takenao; Kai, Tetsuya; Oikawa, Kenichi
Journal of Nuclear Materials, 569, p.153921_1 - 153921_6, 2022/10
Times Cited Count:1 Percentile:45.11(Materials Science, Multidisciplinary)Akiyama, Daisuke*; Kusaka, Ryoji; Kumagai, Yuta; Nakada, Masami; Watanabe, Masayuki; Okamoto, Yoshihiro; Nagai, Takayuki; Sato, Nobuaki*; Kirishima, Akira*
Journal of Nuclear Materials, 568, p.153847_1 - 153847_10, 2022/09
Times Cited Count:2 Percentile:68.58(Materials Science, Multidisciplinary)FeUO, CrUO
, and Fe
Cr
UO
are monouranates containing pentavalent U. Even though these compounds have similar crystal structures, their formation conditions and thermal stability are significantly different. To determine the factors causing the difference in thermal stability between FeUO
and CrUO
, their crystal structures were evaluated in detail. A Raman band was observed at 700 cm
in all the samples. This Raman band was derived from the stretching vibration of the O-U-O axis band, indicating that Fe
Cr
UO
was composed of a uranyl-like structure in its lattice regardless of its "x"' value. M
ssbauer measurements indicated that the Fe in FeUO
and Fe
Cr
UO
were trivalent. Furthermore, Fe
Cr
UO
lost its symmetry around Fe
with increasing electron densities around Fe
, as the abundance of Cr increased. These results suggested no significant structural differences between FeUO
and CrUO
. Thermogravimetric measurements for UO
, FeUO
, and CrUO
showed that the temperature at which FeUO
decomposed under an oxidizing condition (approximately 800
C) was significantly lower than the temperature at which the decomposition of CrUO
started (approximately 1250
C). Based on these results, we concluded that the decomposition of FeUO
was triggered by an "in-crystal" redox reaction, i.e., Fe
U
Fe
U
, which would not occur in the CrUO
lattice because Cr
could never be reduced under the investigated condition. Finally, the existence of Cr
in FexCr
UO
effectively suppressed the decomposition of the Fe
Cr
UO
crystal, even at a very low Cr content.
Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Journal of Nuclear Materials, 568, p.153865_1 - 153865_12, 2022/09
Times Cited Count:3 Percentile:81.34(Materials Science, Multidisciplinary)The normal spectral emissivity, specific heat capacity and thermal conductivity of type 316 austenitic stainless steel (SS) containing boron carbide (BC) in a liquid state were experimentally measured over the composition range of SS-
mass% B
C (up to 10%) and wide temperature ranges using an electromagnetic levitator in a static magnetic field. The normal spectral emissivity and specific heat capacity were almost constant against temperature for all SS-B
C melts, and the thermal conductivities of the melts had a negligible or small positive temperature dependence. The B
C-content dependence of each property at 1800 K had a different tendency across the eutectic composition (around 3 mass% B
C) of the SS-B
C pseudo-binary system.
Kirishima, Akira*; Akiyama, Daisuke*; Kumagai, Yuta; Kusaka, Ryoji; Nakada, Masami; Watanabe, Masayuki; Sasaki, Takayuki*; Sato, Nobuaki*
Journal of Nuclear Materials, 567, p.153842_1 - 153842_15, 2022/08
Times Cited Count:3 Percentile:81.34(Materials Science, Multidisciplinary)To understand the chemical structure and stability of nuclear fuel debris consisting of UO, Zr, and Stainless Steel (SUS) generated by the Fukushima Daiichi Nuclear Power Plant accident in Japan in 2011, simulated debris of the UO
-SUS-Zr system and other fundamental component systems were synthesized and characterized. The simulated debris were synthesized by heat treatment for 1 to 12 h at 1600
C, in inert (Ar) or oxidative (Ar + 2% O
) atmospheres.
Np and
Am tracers were doped for the leaching tests of these elements and U from the simulated debris. The characterization of the simulated debris was conducted by XRD, SEM-EDX, Raman spectroscopy, and M
ssbauer spectroscopy, which provided the major uranium phase of the UO
-SUS-Zr debris was the solid solution of U
O
(s.s.) with Zr(IV) and Fe(II) regardless of the treatment atmosphere. The long-term immersion test of the simulated debris in pure water and that in seawater revealed the macro scale crystal structure of the simulated debris was chemically very stable in the wet condition for a year or more. Furthermore, the leaching test results showed that the actinide leaching ratios of U, Np, Am from the UO
-SUS-Zr debris were very limited and less than 0.08 % for all the experiments in this study.
Shirasaki, Kenji*; Tabata, Chihiro*; Sunaga, Ayaki*; Sakai, Hironori; Li, D.*; Konaka, Mariko*; Yamamura, Tomoo*
Journal of Nuclear Materials, 563, p.153608_1 - 153608_11, 2022/05
Times Cited Count:2 Percentile:68.58(Materials Science, Multidisciplinary)We focused on the direct synthesis of (U, )O
solid solution (
=Th, Np) by extending our recent progress in hydrothermal synthesis with additives. The homogeneity of the (U,
)O
(
= Th, Np) systems prepared by supercritical hydrothermal reactions was investigated through crystallographic analysis based on Vegard's law, and the
Na nuclear magnetic resonance (NMR) measurement of (U, Np, Na)O
solid solutions. Our experimental and analytical results revealed that (i) an optimal additive is ammonium carbonate and starting uranium valence is IV in the case of (U, Th)O
, and (ii) an optimal additive is ethanol and starting uranium valence is VI in the case of (U, Np)O
, for producing the homogeneous solid solutions by hydrothermal synthesis.
Uehara, Akihiro*; Akiyama, Daisuke*; Ikeda, Atsushi; Numako, Chiya*; Terada, Yasuko*; Nitta, Kiyofumi*; Ina, Toshiaki*; Takeda-Homma, Shino*; Kirishima, Akira*; Sato, Nobuaki*
Journal of Nuclear Materials, 559, p.153422_1 - 153422_11, 2022/02
Times Cited Count:2 Percentile:68.58(Materials Science, Multidisciplinary)Wang, J.*; Hatano, Yuji*; Toyama, Takeshi*; Suzudo, Tomoaki; Hinoki, Tatsuya*; Alimov, V. Kh.*; Schwarz-Selinger, T.*
Journal of Nuclear Materials, 559, p.153449_1 - 153449_7, 2022/02
Times Cited Count:2 Percentile:68.58(Materials Science, Multidisciplinary)To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523-1273 K. These displacement-damaged samples were exposed to D gas at a temperature of 673 K. The addition of 0.3% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature. Positron lifetime for W-0.3Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects by 0.3% Cr addition.