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Journal Articles

Spatial distribution and preferred orientation of crystalline microstructure of lead-bismuth eutectic

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

Journal Articles

Study on the relation between the crystal structure and thermal stability of FeUO$$_{4}$$ and CrUO$$_{4}$$

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

FeUO$$_{4}$$, CrUO$$_{4}$$, and Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ 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$$_{4}$$ and CrUO$$_{4}$$, their crystal structures were evaluated in detail. A Raman band was observed at 700 cm$$^{-1}$$ in all the samples. This Raman band was derived from the stretching vibration of the O-U-O axis band, indicating that Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ was composed of a uranyl-like structure in its lattice regardless of its "x"' value. M$"o$ssbauer measurements indicated that the Fe in FeUO$$_{4}$$ and Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ were trivalent. Furthermore, Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ lost its symmetry around Fe$$^{mathrm{III}}$$ with increasing electron densities around Fe$$^{mathrm{III}}$$, as the abundance of Cr increased. These results suggested no significant structural differences between FeUO$$_{4}$$ and CrUO$$_{4}$$. Thermogravimetric measurements for UO$$_{2}$$, FeUO$$_{4}$$, and CrUO$$_{4}$$ showed that the temperature at which FeUO$$_{4}$$ decomposed under an oxidizing condition (approximately 800 $$^{circ}$$C) was significantly lower than the temperature at which the decomposition of CrUO$$_{4}$$ started (approximately 1250 $$^{circ}$$C). Based on these results, we concluded that the decomposition of FeUO$$_{4}$$ was triggered by an ``in-crystal'' redox reaction, i.e., Fe$$^{mathrm{III}}$$ $${+}$$ U$$^{mathrm{V}}$$ $$rightarrow$$ Fe$$^{mathrm{II}}$$ $${+}$$ U$$^{mathrm{VI}}$$, which would not occur in the CrUO$$_{4}$$ lattice because Cr$$^{mathrm{III}}$$ could never be reduced under the investigated condition. Finally, the existence of Cr$$^{mathrm{III}}$$ in FexCr$$_{1-x}$$UO$$_{4}$$ effectively suppressed the decomposition of the Fe$$_{x}$$Cr$$_{1-x}$$UO$$_{4}$$ crystal, even at a very low Cr content.

Journal Articles

Normal spectral emissivity, specific heat capacity, and thermal conductivity of type 316 austenitic stainless steel containing up to 10 mass% B$$_{4}$$C in a liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Journal of Nuclear Materials, 568, p.153865_1 - 153865_12, 2022/09

 Times Cited Count:0 Percentile:0.02(Materials Science, Multidisciplinary)

The normal spectral emissivity, specific heat capacity and thermal conductivity of type 316 austenitic stainless steel (SS) containing boron carbide (B$$_{4}$$C) in a liquid state were experimentally measured over the composition range of SS-$$x$$ mass% B$$_{4}$$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$$_{4}$$C melts, and the thermal conductivities of the melts had a negligible or small positive temperature dependence. The B$$_{4}$$C-content dependence of each property at 1800 K had a different tendency across the eutectic composition (around 3 mass% B$$_{4}$$C) of the SS-B$$_{4}$$C pseudo-binary system.

Journal Articles

Structure, stability, and actinide leaching of simulated nuclear fuel debris synthesized from UO$$_{2}$$, Zr, and stainless-steel

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:0 Percentile:0.02(Materials Science, Multidisciplinary)

To understand the chemical structure and stability of nuclear fuel debris consisting of UO$$_{2}$$, Zr, and Stainless Steel (SUS) generated by the Fukushima Daiichi Nuclear Power Plant accident in Japan in 2011, simulated debris of the UO$$_{2}$$-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$$^{circ}$$C, in inert (Ar) or oxidative (Ar + 2% O$$_{2}$$) atmospheres. $$^{237}$$Np and $$^{241}$$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$"o$ssbauer spectroscopy, which provided the major uranium phase of the UO $$_{2}$$-SUS-Zr debris was the solid solution of U$$^{mathrm{IV}}$$O$$_{2}$$ (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$$_{2}$$-SUS-Zr debris were very limited and less than 0.08 % for all the experiments in this study.

Journal Articles

Homogeneity of (U, $$M$$)O$$_2$$ ($$M$$ = Th, Np) prepared by supercritical hydrothermal synthesis

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:1 Percentile:87.44(Materials Science, Multidisciplinary)

We focused on the direct synthesis of (U, $$M$$)O$$_2$$ solid solution ($$M$$=Th, Np) by extending our recent progress in hydrothermal synthesis with additives. The homogeneity of the (U, $$M$$)O$$_2$$ ($$M$$ = Th, Np) systems prepared by supercritical hydrothermal reactions was investigated through crystallographic analysis based on Vegard's law, and the $$^{23}$$Na nuclear magnetic resonance (NMR) measurement of (U, Np, Na)O$$_2$$ 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$$_{2+x}$$, and (ii) an optimal additive is ethanol and starting uranium valence is VI in the case of (U, Np)O$$_{2+x}$$, for producing the homogeneous solid solutions by hydrothermal synthesis.

Journal Articles

Speciation on the reaction of uranium and zirconium oxides treated under oxidizing and reducing atmospheres

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:1 Percentile:87.44(Materials Science, Multidisciplinary)

Journal Articles

Suppression of vacancy formation and hydrogen isotope retention in irradiated tungsten by addition of chromium

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:0 Percentile:0.02(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$$_{2}$$ 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.

Journal Articles

Electrochemical recovery of Zr and Cd from molten chloride salts for reprocessing of used nitride fuels

Murakami, Tsuyoshi*; Hayashi, Hirokazu

Journal of Nuclear Materials, 558, p.153330_1 - 153330_7, 2022/01

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

Excess amounts of dissolution agents, CdCl$$_2$$ and ZrCl$$_4$$, are required to dissolve transuranium (TRU: Pu and minor actinides) nitrides into LiCl-KCl melts at the chemical dissolution step, which is the first step in the reprocessing of used nitride fuels. We propose an electrochemical process where the remaining Zr and Cd are recovered from the melts to be recycled as dissolution agents for the chemical dissolution step, leaving TRU in the melts. Since the initial concentration ratio of CdCl$$_2$$/ZrCl$$_4$$ remaining in the melts would depend on the condition of the chemical dissolution step and would vary during the proposed electrochemical recovery process, electrochemical behaviors of Zr and Cd were investigated in LiCl-KCl melts with various concentration ratios of CdCl$$_2$$/ZrCl$$_4$$ at 723 K to confirm the basic feasibility of the proposed process. Potentiostatic electrolysis was performed using a liquid Cd cathode at -1.05 V (vs. Ag/AgCl), which was a more positive potential than the redox potentials of TRU on the liquid Cd electrode. The obtained results showed that the current efficiency for recovering Zr and Cd from the melts was as high as 100% regardless of the CdCl$$_2$$/ZrCl$$_4$$ concentration ratio in the melts.

Journal Articles

Relative oxygen potential measurements of (U,Pu)O$$_{2}$$ with Pu = 0.45 and 0.68 and related defect formation energy

Hiroka, Shun; Matsumoto, Taku; Sunaoshi, Takeo*; Hino, Tetsushi*

Journal of Nuclear Materials, 558, p.153375_1 - 153375_8, 2022/01

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Structural characterization by X-ray analytical techniques of calcium aluminate cement modified with sodium polyphosphate containing cesium chloride

Takahatake, Yoko; Watanabe, So; Irisawa, Keita; Shiwaku, Hideaki; Watanabe, Masayuki

Journal of Nuclear Materials, 556, p.153170_1 - 153170_7, 2021/12

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

Journal Articles

The Role of silicon on solute clustering and embrittlement in highly neutron-irradiated pressurized water reactor surveillance test specimens

Takamizawa, Hisashi; Hata, Kuniki; Nishiyama, Yutaka; Toyama, Takeshi*; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 556, p.153203_1 - 153203_10, 2021/12

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

Solute clusters (SCs) formed in pressurized water reactor surveillance test specimens neutron-irradiated to a fluence of 1 $$times$$ 10$$^{20}$$ n/cm$$^{2}$$ were analyzed via atom probe tomography to understand the effect of silicon on solute clustering and irradiation embrittlement of reactor pressure vessel steels. In high-Cu bearing materials, Cu atoms were aggregated at the center of cluster surrounded by the Ni, Mn, and Si atoms like a core-shell structure. In low-Cu bearing materials, Ni, Mn, and Si atoms formed cluster and these solutes were not comprised core-shell structure in SCs. While the number of Cu atoms in clusters was decreased with decreasing nominal Cu content, the number of Si atoms had clearly increased. The cluster radius ($$r$$) and number density ($$N_{d}$$) decreased and increased, respectively, with increasing nominal Si content. The shift in the reference temperature for nil-ductility transition ($$Delta$$RT$$_{NDT}$$) showed a good correlation with the square root of volume fraction ($$V_{f}$$) multiplied by r ($$sqrt{V_{f}times {r}}$$). This suggested that the dislocation cutting through the particles mechanism dominates the precipitation hardening responsible for irradiation embrittlement. The negative relation between the nominal Si content and $$Delta$$RT$$_{NDT}$$ indicated that increasing of nominal Si content reduces the degree of embrittlement.

Journal Articles

Radiation-enhanced diffusion of copper in iron studied by three-dimensional atom probe

Toyama, Takeshi*; Suzudo, Tomoaki; Nagai, Yasuyoshi*; 9 of others*

Journal of Nuclear Materials, 556, p.153176_1 - 153176_7, 2021/12

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

We performed a high-precision investigation of radiation-enhanced diffusion (RED) using electron irradiation and three-dimensional atom probe (3D-AP). Cu-Fe diffusion pairs were created using high-purity Fe and Cu as base materials, and irradiated by 2 MeV electron. Cu diffusion into the Fe matrix was observed at the atomic level using 3D-AP, and the diffusion coefficient was obtained directly using Fick's law. RED was clearly observed, and the ratio of diffusion under irradiation to thermal diffusion was enhanced at low temperature. RED was quantitatively evaluated using the reaction kinetics model, and the model which consider only vacancies gave a good agreement. This gave experimental clarification that RED was dominated by irradiation-induced vacancies. In addition, the direct experimental results on the effect of irradiation on the solubility limits of Cu in Fe was obtained; solubility limits under irradiation were found to be lower than those under thermal aging.

Journal Articles

Evaluation of the dissolution behavior of zircon using high-resolution phase-shift interferometry microscope

Kitagaki, Toru

Journal of Nuclear Materials, 557, p.153254_1 - 153254_8, 2021/12

 Times Cited Count:1 Percentile:44.81(Materials Science, Multidisciplinary)

Journal Articles

Tensile properties on dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging

Yano, Yasuhide; Tanno, Takashi; Oka, Hiroshi; Otsuka, Satoshi; Kaito, Takeji

Journal of Nuclear Materials, 555, p.153105_1 - 153105_8, 2021/11

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

The aim of this study was to evaluate the tensile properties and microstructures of dissimilar welds between 11Cr-ferritic/martensitic steel and 316 stainless steel after thermal aging at temperatures between 400 and 600$$^{circ}$$C up to 30,000 h. Characterization of microstructure was carried out by scanning electron microscopy and transmission electron microscopy. Microstructural analysis showed that the microstructure in the weld metals consisted of lath martensite containing a small amount of residual austenite. Thermal aging hardening of WMs occurred at 400 and 450$$^{circ}$$C due to the effects of both a-a' phase separation and G-phase precipitation. However, there was no significant change in the total elongation, and fracture surfaces indicated that very fine dimpled rupture was predominant rather than the cleavage rupture. It was suggested that lath martensite phases enhanced the tensile strength due to phase separation, while residual austenite played a role in keeping elongation as a soft phase.

Journal Articles

Effect of B$$_{4}$$C addition on the solidus and liquidus temperatures, density and surface tension of type 316 austenitic stainless steel in the liquid state

Fukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa

Journal of Nuclear Materials, 554, p.153100_1 - 153100_11, 2021/10

 Times Cited Count:3 Percentile:81.44(Materials Science, Multidisciplinary)

The effects of B$$_{4}$$C addition on the solidus and liquidus temperatures of type 316 austenitic stainless steel (SS), and on the density and surface tension of molten SS, were experimentally studied. The solidus temperature of SS-x mass% B$$_{4}$$C (from 0 to 10) monotonically decreased from 1666 to 1307 K with B$$_{4}$$C addition. The liquidus temperature had a minimum at around 2.5 mass% B$$_{4}$$C, and increased with further B$$_{4}$$C addition up to 10 mass%. The density and surface tension of molten SS-x mass %B$$_{4}$$C were successfully measured over a wide temperature range (including an undercooling region) via an electromagnetic-levitation technique. The density of each sample decreased linearly with temperature. The density also monotonically decreased with B$$_{4}$$C content. Although the addition of B$$_{4}$$C had no clear effect on the surface tension of SS-x mass %B$$_{4}$$C, sulfur dissolved in SS316L caused a significant decrease in the surface tension.

Journal Articles

Perspectives on multiscale modelling and experiments to accelerate materials development for fusion

Gilbert, M. R.*; Arakawa, Kazuto*; Suzudo, Tomoaki; Tsuru, Tomohito; 26 of others*

Journal of Nuclear Materials, 554, p.153113_1 - 153113_31, 2021/10

 Times Cited Count:10 Percentile:93.13(Materials Science, Multidisciplinary)

Modelling will continue to do so until the first generation of fusion power plants come on line and allow long-term behaviour to be observed. In the meantime, the modelling is supported by experiments that attempt to replicate some aspects of the eventual operational conditions. In 2019, a group of leading experts met under the umbrella of the IEA to discuss the current position and ongoing challenges in modelling of fusion materials and how advanced experimental characterisation is aiding model improvement. This review draws from the discussions held during that workshop. Topics covering modelling of irradiation-induced defect production and fundamental properties, gas behaviour, clustering and segregation, defect evolution and interactions are discussed, as well as new and novel multiscale simulation approaches, and the latest efforts to link modelling to experiments through advanced observation and characterisation techniques.

Journal Articles

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hiroka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

 Times Cited Count:2 Percentile:68.98(Materials Science, Multidisciplinary)

To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.

Journal Articles

Viscosities of molten B$$_{4}$$C-stainless steel alloys

Nishi, Tsuyoshi*; Sato, Rika*; Ota, Hiromichi*; Kokubo, Hiroki*; Yamano, Hidemasa

Journal of Nuclear Materials, 552, p.153002_1 - 153002_7, 2021/08

 Times Cited Count:2 Percentile:68.98(Materials Science, Multidisciplinary)

Determining high precision viscosities of molten B$$_{4}$$C-stainless steel (B$$_{4}$$C-SS) alloys is essential for the core disruptive accident analyses of sodium-cooled fast reactors and for analysis of severe accidents in boiling water reactors (BWR) as appeared in Fukushima Daiichi. However, there are no data on the high precision viscosities of molten B$$_{4}$$C-SS alloys due to experimental difficulties. In this study, the viscosities of molten SS (Type 316L), 2.5mass%B$$_{4}$$C-SS, 5.0mass%B$$_{4}$$C-SS, and 7.0mass%B$$_{4}$$C-SS alloys were measured using the oscillating crucible method in temperature ranges of 1693-1793 K, 1613-1793 K, 1613-1793 K, and 1713-1793 K, respectively. The viscosity was observed to increase as the B$$_{4}$$C concentration increased from 0 to 7.0 mass%. Using the experimental data of the molten 2.5mass%B$$_{4}$$C-SS and 5.0mass%B$$_{4}$$C-SS and 7.0mass%B$$_{4}$$C-SS in the temperature range of 1713-1793 K, the equation for the viscosity of molten B$$_{4}$$C-SS alloys was determined, and the measurement error of the viscosity of molten B$$_{4}$$C-SS alloys is less than $$pm$$8%.

Journal Articles

Microstructural stability of ODS steel after very long-term creep test

Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki

Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04

 Times Cited Count:3 Percentile:81.44(Materials Science, Multidisciplinary)

In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700$$^{circ}$$C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.

Journal Articles

Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels; Using nickel isotopes to simulate fusion neutron response

Kim, B. K.*; Tan, L.*; Sakasegawa, Hideo; Parish, C. M.*; Zhong, W.*; Tanigawa, Hiroyasu*; Kato, Yutai*

Journal of Nuclear Materials, 545, p.152634_1 - 152634_12, 2021/03

 Times Cited Count:1 Percentile:44.81(Materials Science, Multidisciplinary)

1414 (Records 1-20 displayed on this page)