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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

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

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:0.02(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:0 Percentile:0.02(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

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:0 Percentile:0.02(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:0 Percentile:0.02(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:1 Percentile:0.02(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

Grain-boundary phosphorus segregation in highly neutron-irradiated reactor pressure vessel steels and its effect on irradiation embrittlement

Hata, Kuniki; Takamizawa, Hisashi; Hojo, Tomohiro*; Ebihara, Kenichi; Nishiyama, Yutaka; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 543, p.152564_1 - 152564_10, 2021/01

 Times Cited Count:1 Percentile:83.53(Materials Science, Multidisciplinary)

Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.2$$times$$10$$^{20}$$ n/cm$$^{2}$$ (E $$>$$ 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation was analyzed using Auger electron spectroscopy (AES) on intergranular facets and found to increase with increasing neutron fluence. A rate theory model was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.% (the maximum P concentration found in RPV steels used in Japanese nuclear power plants intended for restart) was estimated to be less than 0.1 in monolayer coverage at 1.0$$times$$10$$^{20}$$ n/cm$$^{2}$$ (E $$>$$ 1 MeV). A comparison of the PWR data with the MTR data showed that neutron flux had no effect upon grain-boundary P segregation. The effects of grain-boundary P segregation upon changes in irradiation hardening and ductile-brittle transition temperature (DBTT) shifts were also discussed. A linear relationship between irradiation hardening and the DBTT shift with a slope of 0.63 obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs.

Journal Articles

Solidification and re-melting mechanisms of SUS-B$$_{4}$$C eutectic mixture

Sumita, Takehiro; Kitagaki, Toru; Takano, Masahide; Ikeda, Atsushi

Journal of Nuclear Materials, 543, p.152527_1 - 152527_15, 2021/01

 Times Cited Count:1 Percentile:0.02(Materials Science, Multidisciplinary)

Journal Articles

Oxygen potential measurement of (U,Pu,Am)O$$_{2 pm x}$$ and (U,Pu,Am,Np)O$$_{2 pm x}$$

Hiroka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

 Times Cited Count:1 Percentile:42.23(Materials Science, Multidisciplinary)

The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (U$$_{0.623}$$Pu$$_{0.350}$$Am$$_{0.027}$$)O$$_{2}$$ and at 1,873 and 1,923 K for (U$$_{0.553}$$Pu$$_{0.285}$$Am$$_{0.015}$$Np$$_{0.147}$$)O$$_{2}$$ by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO$$_{2}$$ and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.

Journal Articles

Oxygen self-diffusion in near stoichiometric (U,Pu)O$$_{2}$$ at high temperatures of 1673-1873 K

Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*

Journal of Nuclear Materials, 542, p.152472_1 - 152472_7, 2020/12

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

The oxygen self-diffusion coefficients in near stoichiometric (U,Pu)O$$_{2}$$ at high temperatures were successfully measured by thermogravimetry combined with the oxygen isotope exchange method. The activation energy for oxygen diffusion in the stoichiometric composition of (U,Pu)O$$_{2}$$ was evaluated from experimental data, and the value was determined to be 248 kJ/mol. In addition, the defect migration energies of (U,Pu)O$$_{2 pm x}$$ were derived, and the oxygen self-diffusion coefficients were evaluated using these. As a result, good agreement was found between the experimental data and the oxygen self-diffusion coefficients calculated using the defect migration energies.

Journal Articles

Atomistic modeling of hardening in spinodally-decomposed Fe-Cr binary alloys

Suzudo, Tomoaki; Takamizawa, Hisashi; Nishiyama, Yutaka; Caro, A.*; Toyama, Takeshi*; Nagai, Yasuyoshi*

Journal of Nuclear Materials, 540, p.152306_1 - 152306_10, 2020/11

 Times Cited Count:1 Percentile:42.23(Materials Science, Multidisciplinary)

Spinodal decomposition in thermally aged Fe-Cr alloys leads to significant hardening, which is the direct cause of the so-called 475C-embrittlement. To illustrate how spinodal decomposition induces hardening by atomistic interactions, we conducted a series of numerical simulations as well as reference experiments. The numerical results indicated that the hardness scales linearly with the short-range order (SRO) parameter, while the experimental result reproduced this relationship within statistical error. Both seemingly suggest that neighboring Cr-Cr atomic pairs essentially cause hardening, because SRO is by definition uniquely dependent on the appearance probability of such pairs. A further numerical investigation supported this notion, as it suggests that the dominant cause of hardening is the pinning effect of dislocations passing over such Cr-Cr pairs.

Journal Articles

Estimation of reliable displacements-per-atom based on athermal-recombination-corrected model in radiation environments at nuclear fission, fusion, and accelerator facilities

Iwamoto, Yosuke; Meigo, Shinichiro; Hashimoto, Shintaro

Journal of Nuclear Materials, 538, p.152261_1 - 152261_9, 2020/09

 Times Cited Count:2 Percentile:66.08(Materials Science, Multidisciplinary)

The displacements-per-atom (dpa) is widely used as an exposure unit to predict the operating lifetime of materials in radiation environments. Because the athermal-recombination-corrected dpa (arc-dpa) model is a more realistic model than the standard Norgertt-Robinson-Torrens (NRT) model, new evaluation of radiation damage will be performed using the arc-dpa model as a standard. In this work, the recent arc-dpa model of various materials are incorporated in PHITS, and the rescaling factors (NRT-dpa/arc-dpa) over a wide energy range are reported. For neutron incidences with the energy spectrum determined in selected nuclear facilities and proton incidences with energies of 600 MeV-50 GeV, the rescaling factor for each material is independent of these irradiation conditions with almost the same value for each material. Our findings will be beneficial for rescaling the NRT-dpa model used for radiation damage applications over a wide energy region.

Journal Articles

Release behavior of radionuclides from MOX fuels irradiated in a fast reactor during heating tests

Tanaka, Kosuke; Sato, Isamu*; Onishi, Takashi; Ishikawa, Takashi; Hirosawa, Takashi; Katsuyama, Kozo; Seino, Hiroshi; Ohno, Shuji; Hamada, Hirotsugu; Tokoro, Daishiro*; et al.

Journal of Nuclear Materials, 536, p.152119_1 - 152119_8, 2020/08

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

In order to obtain the release rate coefficients from fuels for fast reactors (FRs), heating tests and the subsequent analyses of the fission products (FPs) and actinides that are released were carried out using samples of uranium-plutonium mixed oxide (MOX) fuel pellets irradiated at the experimental fast reactor Joyo. Three heating tests targeting temperatures of 2773, 2973 and 3173 K were conducted using an FP release behavior test apparatus equipped with a high-frequency induction furnace and solid FP sampling systems consisting of a thermal gradient tube (TGT) and filters. Irradiated fuel pellets were placed into a tungsten crucible, then loaded into the induction furnace. The temperature was raised continuously at a heating rate of 10 K/s to the targeted temperature and maintained for 500 s in a flowing argon gas atmosphere. The FPs and actinides released from the MOX fuels and deposited in the TGT and filters were quantified by gamma-ray spectrometry and inductively coupled plasma mass spectrometry (ICP-MS) analysis. Based on the analysis, the release rates of radionuclides from MOX fuels for FR were obtained and compared with literature data for light water reactor (LWR) fuels. The release rate coefficients of FPs obtained in this study were found to be similar to or lower than the literature values for LWR fuels. It was also found that the release rate coefficient data for actinides were within the range of variation of literature values for LWR fuels.

Journal Articles

Effect of O/M ratio on sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$

Nakamichi, Shinya; Hiroka, Shun; Kato, Masato; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 535, p.152188_1 - 152188_8, 2020/07

 Times Cited Count:3 Percentile:78.97(Materials Science, Multidisciplinary)

Oxygen-to-metal ratio (O/M) of uranium and plutonium mixed oxide depends on its oxygen partial pressure. To attain the desirable microstructure and O/M ratio of sintered pellets, it is important to investigate the relation between the sintering behavior and the atmosphere of sintering process. In this study, sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$ and (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ in controlled po$$_{2}$$ atmosphere were investigated. It was found activation energy of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ was higher than that of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$. On the other hand, it was observed grain growth during sintering was suppressed in hypo-stoichiometric composition.

Journal Articles

Experimental validation of tensile properties measured with thick samples taken from MEGAPIE target

Saito, Shigeru; Suzuki, Kazuhiro; Hatakeyama, Yuichi; Suzuki, Miho; Dai, Y.*

Journal of Nuclear Materials, 534, p.152146_1 - 152146_16, 2020/06

 Times Cited Count:1 Percentile:42.23(Materials Science, Multidisciplinary)

A post-irradiation examination (PIE) was performed on the tensile specimens prepared from the MEGAPIE (MEGAwatt Pilot Experiment) target which were irradiated in flowing lead-bismuth eutectic (LBE). Thicknesses of the specimens were over two times larger than that of the standard specimen. The PIE revealed that the T91 specimens showed a 1.5-2.0 times larger total elongation (TE) compared to the literature values for a specimen with standard t/w (ratio of thickness to width). It could be suggested that the t/w and TE were strongly correlated. Then, we tried to investigate the effects of the t/w on the TE by comparing unirradiated specimens. We found that there was no t/w dependence on the strength and uniform elongation. On the other hand, the TE increases with increasing t/w. Based on the experimental data, we correlated the TE with various specimens t/w to estimate appropriate TE values, including that for the standard specimen.

Journal Articles

Segregation behavior of Fe and Gd in molten corium during solidification progress

Sudo, Ayako; Meszaros, B.*; Poznyak, I.*; Sato, Takumi; Nagae, Yuji; Kurata, Masaki

Journal of Nuclear Materials, 533, p.152093_1 - 152093_8, 2020/05

 Times Cited Count:1 Percentile:42.23(Materials Science, Multidisciplinary)

Journal Articles

Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

Cappia, F.*; Tanaka, Kosuke; Kato, Masato; McClellan, K.*; Harp, J.*

Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05

 Times Cited Count:1 Percentile:42.23(Materials Science, Multidisciplinary)

Journal Articles

Oxidation kinetics of silicon carbide in steam at temperature range of 1400 to 1800$$^{circ}$$C studied by laser heating

Pham, V. H.; Nagae, Yuji; Kurata, Masaki; Bottomley, D.; Furumoto, Kenichiro*

Journal of Nuclear Materials, 529, p.151939_1 - 151939_8, 2020/02

AA2019-0197.pdf:1.61MB

 Times Cited Count:2 Percentile:66.08(Materials Science, Multidisciplinary)

Journal Articles

Anomalous small-angle X-ray scattering (ASAXS) study of irradiation-induced nanostructure change in Fe-ion beam irradiated oxide dispersion-strengthened (ODS) steel

Kumada, Takayuki; Oba, Yojiro; Motokawa, Ryuhei; Morooka, Satoshi; Tominaga, Aki; Tanida, Hajime; Shobu, Takahisa; Konno, Azusa; Owada, Kenji*; Ono, Naoko*; et al.

Journal of Nuclear Materials, 528, p.151890_1 - 151890_7, 2020/01

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

We have developed an anomalous small-angle X-ray scattering (ASAXS) diffractometer in SPring-8 to investigate irradiation-induced nanostructural change in ion-beam irradiated stainless steel. A thermally-aged MA956 stainless steel sample displays power-law scattering that follows the Porod law at the magnitude of scattering vector, Q, below 0.5 nm$$^-1$$ and an overlapped shoulder around 0.7 nm$$^-1$$. After the ion-beam irradiation, the intensity of the shoulder remained unchanged, whereas that of the power-law scattering nearly doubled. This result indicates that none of the structural parameters of the Cr-rich nanoprecipitates, such as the number density, size, and interface roughness, were changed by the irradiation.

Journal Articles

Material characterization of the VULCANO corium concrete interaction test with concrete representative of Fukushima Daiichi Nuclear Plants

Brissonneau, L.*; Ikeuchi, Hirotomo; Piluso, P.*; Gousseau, J.*; David, C.*; Testud, V.*; Roger, J.*; Bouyer, V.*; Kitagaki, Toru; Nakayoshi, Akira; et al.

Journal of Nuclear Materials, 528, p.151860_1 - 151860_18, 2020/01

 Times Cited Count:5 Percentile:86.53(Materials Science, Multidisciplinary)

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