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Journal Articles

An Investigation on the control rod homogenization method for next-generation fast reactor cores

Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo

Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

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

Comparison of sodium fast reactor core assembly seismic evaluation using the Japanese and French simulation tools

Yamamoto, Tomohiko; Matsubara, Shinichiro*; Harada, Hidenori*; Saunier, P.*; Martin, L.*; Gentet, D.*; Dirat, J.-F.*; Collignon, C.*

Nuclear Engineering and Design, 383, p.111406_1 - 111406_14, 2021/11

Japan-France collaboration on ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project is launched in 2014. In this project, Japan-France evaluates core assemblies with interferences on seismic event. The object of this study is to verify the seismic evaluation method on core assemblies between Japan and France by comparing the results. The analysis of this benchmark calculation shows a satisfactory agreement between the Japanese and French tools and the figures show a good behavior of the core in horizontal direction under French seismic condition.

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

Initial sintering kinetics of non-stoichiometric CeO$$_{2-x}$$

Watanabe, Masashi; Seki, Takayuki*

Materials Science & Engineering B, 272, p.115369_1 - 115369_6, 2021/10

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

The effect of oxygen non-stoichiometry on the initial sintering behavior of CeO$$_{2}$$ was investigated. It was found that the initial sintering of the stoichiometric and hypo-stoichiometric composition was controlled by the grain boundary diffusion. The activation energies of cation diffusion were derived from initial sintering data. Moreover, it is suggested that the cation diffusion was caused by a vacancy mechanism.

Journal Articles

Wettability of pure metals with liquid sodium and liquid tin

Saito, Junichi; Kobayashi, Yohei*; Shibutani, Hideo*

Materials Transactions, 62(10), p.1524 - 1532, 2021/10

no abstracts in English

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

Laser beam direct energy deposition of graded austenitic-to-martensitic steel junctions compared to dissimilar electron beam welding

Villaret, F.*; Boulnat, X.*; Aubry, P.*; Yano, Yasuhide; Otsuka, Satoshi; Fabregue, D.*; de Carlan, Y.*

Materials Science & Engineering A, 824, p.141794_1 - 141794_10, 2021/09

 Times Cited Count:0

Journal Articles

Droplet generation during spray impact onto a downward-facing solid surface

Zhan, Y.*; Sun, G.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Takata, Takashi

Experimental Thermal and Fluid Science, 126, p.110402_1 - 110402_8, 2021/08

 Times Cited Count:0 Percentile:0.01(Thermodynamics)

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

Effects of thermal aging on the mechanical properties of FeCrAl-ODS alloy claddings

Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*

Materials Transactions, 62(8), p.1239 - 1246, 2021/08

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

The FeCrAl-ODS alloy claddings were manufactured and Vickers hardness, ring tensile tests and TEM observations of these claddings were performed to investigate the effects of thermal aging at 450 $$^{circ}$$C for 5,000 and 15,000 h. The age-hardening of all FeCrAl-ODS alloy cladding was found. In addition, the significant increase in tensile strength was accompanied by much larger loss of ductility. It was suggested that this age-hardening behavior was attributed to the (Ti, Al)-enriched phase ($$beta$$' phase) and the $$alpha$$' phase precipitates (content of Al is $$<$$ 7 wt%). In comparison with FeCrAl-ODS alloys with almost same chemical compositions, there was significant age-hardening in both alloys. However, the extrusion bar with no-recrystallized structures was keeping good ductility. It was suggested that this different behavior of reduction ductility was attributed to the effects of grain boundaries, dislocation densities and specimen preparation direction.

Journal Articles

Velocity distribution in the subchannels of a pin bundle with a wrapping wire; Evaluation of the Reynolds number dependence in a three-pin bundle

Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*

Mechanical Engineering Journal (Internet), 8(4), p.20-00547_1 - 20-00547_11, 2021/08

A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and the mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. In the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.

Journal Articles

Formulation of plastic strain distribution derived from long-distance travel of temperature distribution based on residual stress required for elastic shakedown behavior

Okajima, Satoshi

Mechanical Engineering Journal (Internet), 8(4), p.21-00080_1 - 21-00080_15, 2021/08

It was reported that the long distance travel of temperature distribution causes a new type of thermal ratcheting, even in the absence of primary stress. When the distance of temperature travel is moderate, the accumulation of the plastic strain due to this mechanism is finally saturated. We have found the strong relationship between hoop-membrane distributions of accumulated plastic strain and residual stress in this saturated case. Focusing on this relationship, we have aimed to predict the saturated distribution of the plastic strain based on the residual stress distribution that is required for the elastic shakedown behavior. In this paper, based on classical shell theory, we formulated the plastic strain distribution that brings uniform hoop-membrane stress in the given region. The formulated strain distribution was validated by the comparison with the accumulated plastic strain distribution obtained by finite element analyses using an elastic-perfectly plastic material.

Journal Articles

Kinetic study on eutectic reaction between boron carbide and stainless steel by differential thermal analysis

Kikuchi, Shin; Nakamura, Kinya*; Yamano, Hidemasa

Mechanical Engineering Journal (Internet), 8(4), p.20-00542_1 - 20-00542_13, 2021/08

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) and stainless steel (SS) may take place. Thus, kinetic behavior of B$$_{4}$$C-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of B$$_{4}$$C-SS eutectic melting, the thermal analysis using the pellet type samples of B$$_{4}$$C and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the B$$_{4}$$C-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B$$_{4}$$C-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.

Journal Articles

Thermophysical properties of austenitic stainless steel containing boron carbide in a solid state

Takai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa

Mechanical Engineering Journal (Internet), 8(4), p.20-00540_1 - 20-00540_11, 2021/08

In a core disruptive accident scenario, boron carbide, which is used as a control rod material, may melt below the melting temperature of stainless steel owing to the eutectic reaction with them. The eutectic mixture produced is assumed to extensively relocate in the degraded core, and this behavior plays an important role in significantly reducing the neutronic reactivity. However, these behaviors have never been simulated in previous severe accident analysis. To contribute to the improvement of the core disruptive accident analysis code, the thermophysical properties of the eutectic mixture in the solid state were measured, and regression equations that show the temperature (and boron carbide concentration) dependence are created.

Journal Articles

Numerical modeling of radiation heat transfer from combusting droplets for a sodium fire analysis

Aoyagi, Mitsuhiro; Takata, Takashi; Uno, Masayoshi*

Nuclear Engineering and Design, 380, p.111258_1 - 111258_11, 2021/08

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Journal Articles

Droplet entrainment by high-speed gas jet into a liquid pool

Sugimoto, Taro*; Kaneko, Akiko*; Abe, Yutaka*; Uchibori, Akihiro; Kurihara, Akikazu; Takata, Takashi; Ohshima, Hiroyuki

Nuclear Engineering and Design, 380, p.111306_1 - 111306_11, 2021/08

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

Liquid droplet entrainment by a high-speed gas jet is a key phenomenon for evaluation of sodium-water reaction. In this study, a visualization experiment for liquid droplet entrainment by an air jet in a water pool by using frame-straddling method was carried for development of an entrainment model in a sodium-water reaction analysis code. This experiment successfully provided clear images that captured generation and movement of droplets. Droplet diameter and moving speed were obtained at different locations and gas jet velocities from image processing. The measured data contributes phenomena elucidation and model development.

Journal Articles

Development of an integrated computer code system for analyzing irradiation behaviors of a fast reactor fuel

Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi

Nuclear Technology, 207(8), p.1280 - 1289, 2021/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.

Journal Articles

Fundamental study on scheduling of inspection process for fast reactor plants

Suzuki, Masaaki*; Ito, Mari*; Hashidate, Ryuta; Takahashi, Keita; Yada, Hiroki; Takaya, Shigeru

2020 9th International Congress on Advanced Applied Informatics (IIAI-AAI 2020), p.797 - 801, 2021/07

Journal Articles

Expectations for the GIF in its 20th year

Sagayama, Yutaka

Enerugi Rebyu, 41(8), P. 42, 2021/07

GIF was launched in 2001 to develop of GEN-4 reactors through international cooperation. The Policy Group chair rotates among the US, France, and Japan, and Hideki Kamide, Advisor to MEXT of the JAEA, is currently the Chair. The development goals and reactor concept have been summarized as follows, Establishment of the development goals of economic, security, sustainability and proliferation resistance. Selection of reactor types of GFR, SCWR, SFR, VHTR, LFR and MSR. The major achievement is establishment of the safety design criterion, which will become an international standard for SFR and has been developed under the leadership of Japan. In the future, it is important to establish a new mechanism in the demonstration stage and to develop design standards and criteria for maintenance and servicing for international standards for GEN-4 reactors.

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