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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

Application of 1D-CFD coupling method to unprotected loss of heat sink event in EBR-II focusing on thermal stratification in cold pool

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08

To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.

Journal Articles

Three-dimensional structural analysis for enhancing resilience of next-generation nuclear structures under extremely high temperature conditions

Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 9 Pages, 2022/07

Journal Articles

Response reduction effect of seismic isolation system considering uncertainty parameters for seismic margin assessment

Yamano, Hidemasa; Okamura, Shigeki*

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 9 Pages, 2022/07

Seismic response analyses were conducted for the pipe with and without the seismic isolation system based on the response waveforms. This study performed a fragility analysis by setting uncertainty parameters on the basis of existing studies. The comparison results showed that the seismic isolation technology is effective for the pipe to prevent cliff-edge effects. In other words, the seismic margin for the seismically isolated plant is 1.2 times larger than that of the non-isolated plant. To evaluate the response reduction effect, this study focused on response coefficients of components as uncertainty parameters, which were specified within a physically possible range. Even if the uncertainty is considered, the HCLPF for the isolated plant is nearly twice as high as the non-isolated plant, namely the response reduction effect is still significant for the isolated plant. Therefore, the isolation technology is effective to avoid cliff-edge effects.

Journal Articles

Development plan of failure mitigation technologies for improving resilience of nuclear structures

Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07

Utilizing fracture control, we are developing a technology to suppress the expansion of damage caused by an event that exceeds the design assumption. We made a plan to develop three issues; (1) Technology for mitigating failure consequence at extremely high temperatures, (2) Technology for mitigating failure consequence against excessive earthquakes, and (3) Methodology for improving reactor structure resilience.

Journal Articles

Vibration test and fatigue test for failure probability evaluation method with integrated energy

Kinoshita, Takahiro*; Okamura, Shigeki*; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*

Transactions of 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 7 Pages, 2022/07

The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were damaged by cumulative damage like fatigue damage during seismic ground motion. However, general evaluation method for key components under seismic ground motion has been based on static loads and elastic region of materials. More accurate evaluation method for S-PRA, which can evaluate the failure of key components such as reactor vessels, has been actually required. In this study, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results of fatigue tests.

Journal Articles

Viscosity measurements of molten metal using an improved oscillating crucible method

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

International Journal of Thermophysics, 43(6), p.85_1 - 85_15, 2022/06

 Times Cited Count:0 Percentile:0.01(Thermodynamics)

In this study, we developed a simple viscosity measurement method based on the principle of least squares to derive the period and logarithmic decrement of oscillation. To confirm the reproducibility of the proposed method, the viscosity of molten nickel was measured and found to be in good agreement with those reported in the literature. The measurement error was less than $$pm$$3%. Further, the experimental data showed good reproducibility, and the measurements were obtained with high accuracies using the proposed method.

Journal Articles

Viscosity measurements of Molten casting steels for high temperature using an oscillating crucible viscometer

Takatsuka, Yuriko*; Matsumoto, Saori*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Hori, Ayumi*; Hayashi, Kiichiro*; Yamano, Hidemasa

Jikken Rikigaku, 22(2), p.117 - 119, 2022/06

This study clarified the effect of the viscosities of molten casting steels for high temperature by measuring them using the oscillating crucible method. The casting steels for high temperature samples used for viscosity measurements contained 0, 10, 20, 30, 40, and 50 mass% Ni. Viscosities were evaluated using Roscoe's equation and measured in the temperature range of 1693-1803 K.

Journal Articles

Viscosity measurements of molten Ni-based superalloys by the oscillating crucible method

Matsumoto, Saori*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Hori, Ayumi*; Hayashi, Kiichiro*; Yamano, Hidemasa

Jikken Rikigaku, 22(2), p.112 - 116, 2022/06

In this study, the viscosity of the molten Ni-based superalloys were measured by the oscillating crucible method. When the logarithm of the viscosity was plotted against the inverse temperature (Arrhenius-type plot), all alloys showed a good Arrhenius-type linearity over the temperature range investigated. It was also found that the viscosities of molten Ni-based superalloys decreased with increasing the Co contents.

Journal Articles

Activities of the GIF safety and operation project of the sodium-cooled fast reactors

Yamano, Hidemasa; Tsige-Tamirat, H.*; Kang, S. H.*; Summer, T.*; Chenaud, M. S.*; Rozhikhin, E.*; Wang, J.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports various activities in 2019 within the SO project.

Journal Articles

France-Japan collaboration on thermodynamic and kinetic studies of core material mixture in severe accidents of sodium-cooled fast reactors

Quaini, A.*; Goss$'e$, S.*; Payot, F.*; Suteau, C.*; Delacroix, J.*; Saas, L.*; Gubernatis, P.*; Martin-Lopez, E.*; Yamano, Hidemasa; Takai, Toshihide; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04

CEA and JAEA defined new sub-tasks under the current implementing arrangement: Kinetics of interaction in core material mixtures- Physical properties of core material mixtures, High temperature thermodynamic data for the UO$$_{2}$$-Fe-B$$_{4}$$C system, Experimental studies on B$$_{4}$$C-SS kinetics and B$$_{4}$$C-SS eutectic material relocation (freezing), B$$_{4}$$C/SS eutectic and kinetics models for SIMMER code systems, Methodology for the modelling of mixtures liquefaction kinetics. The paper describes major R&D results obtained in the France-Japan collaboration under the previous implementing arrangement as well as experimental and analytical roadmaps under the current arrangement.

Journal Articles

France-Japan collaboration on the SFR severe accident studies; Outcomes and future work program

Kubo, Shigenobu; Payot, F.*; Yamano, Hidemasa; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; Shibata, Akihiro*; et al.

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

Journal Articles

Development of methodology to evaluate mechanical consequences of vapor expansion in SFR severe accident transients; Lessons learned from previous France-Japan collaboration and future objectives and milestones

Bachrata, A.*; Gentet, D.*; Bertrand, F.*; Marie, N.*; Kubota, Ryuzaburo*; Sogabe, Joji; Sasaki, Keisuke; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in the paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents. For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. A fluid structure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In the paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Even if the most penalizing case was evidenced in this paper, no significant RV deformation was observed in both EUROPLEXUS and AUTODYN calculation results. The assumed mechanical energy was small for the core expansion phase.

Journal Articles

Preliminary application of eutectic reaction model on boron carbide and stainless steel to severe accident simulation of sodium-cooled fast reactors

Yamano, Hidemasa; Morita, Koji*

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

For a severe accident (SA) simulation of sodium-cooled fast reactors, a eutectic reaction model between boron carbide (B$$_{4}$$C) and stainless steel (SS) has been developed to be incorporated into the SA simulation codes: SIMMER-III/IV. To confirm the applicability of SIMMER-IV involving the eutectic reaction model to reactor simulations, this study has preliminarily applied this code with the newly developed physical model to a SA simulation of a large-scale SFR designed in Japan. The simulation results show that the eutectic reaction is caused by the contact between the liquid SS and the broken B$$_{4}$$C pellets which are released to the coolant channel after the failure of cladding which is melted by the mixture of liquid SS and fuel particles coming from the neighboring fuel assemblies. The liquid eutectic material formed by the reaction stayed in the control assembly and the neighboring fuel assemblies. This preliminary simulation shows that the spreading area of B$$_{4}$$C-SS eutectic formation is limited within this calculation time.

Journal Articles

Latest trends of advanced reactor development supporting nuclear innovation, 1; Latest trends of advanced reactor development in Japan and foreign countries

Yamano, Hidemasa; Inaba, Yoshitomo

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 64(2), p.74 - 78, 2022/02

This report mainly introduces trends in innovative reactor development related to nuclear innovation toward 2050 carbon neutral in Japan in addition to introducing overseas development trends for major developing countries.

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

Development of effectiveness evaluations technology of the measures for improving resilience of nuclear structures against excessive earthquake

Nishino, Hiroyuki; Onoda, Yuichi; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 10 Pages, 2021/10

The objective of this study is to develop an effectiveness evaluations technology of the measures for improving resilience of nuclear structures against excessive earthquake by introducing the fracture control concept. After analyzing event tree in previous studies of PRA against earthquake, this study identified sequences of protected loss of heat sink and loss of reactor level induced from excessive earthquake as accident sequences in which improving resilience of nuclear structures become effective. This study focused on important components for safety (e.g., reactor vessel, air coolers, pipes of primary loops in decay heat removal systems, etc.) to be used as countermeasures for improving the resilience. Core damage frequency is selected as an index in evaluating effectiveness of the measures for improving the resilience. Seismic safety margin of the components is assumed to be enlarged when the measures for improving the resilience with the fracture control concept are implemented. Through the trial calculation, reduction effect of the core damage frequency was quantified. The result showed that the measures for improving the resilience are significantly effective for decreasing the core damage frequency in excessive earthquake twice higher than a design basis ground motion. The general concept for the effectiveness evaluations technology was formulated.

Journal Articles

Development of effectiveness evaluations technology of the measures for improving resilience of nuclear structures at ultra high temperature

Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

Proceedings of Asian Symposium on Risk Assessment and Management 2021 (ASRAM 2021) (Internet), 11 Pages, 2021/10

The effectiveness evaluations technology of the measures for improving resilience by applying a fracture control concept under ultra-high temperature conditions has developed for prototype sodium-cooled fast reactor Monju as a model plant, and the trial evaluation has conducted using this technology in this paper. The important accident sequences to which the fracture control concept is expected to be applied under ultra-high temperature condition are identified by investigating the results of the existing researches of level-2 probabilistic risk assessment for Monju. Accident sequences categorized in protected loss of heat sink and loss of reactor level are both identified as such important accident sequences which has the potential to prevent core damage. This study has developed the technology to evaluate the effectiveness of improving resilience, where the headings which stand for success or failure of the measures to improve resilience are introduced into the event tree, the branch probability of them is set, and the effectiveness of improving resilience is expressed as the reduction of core damage frequency. As a result of the trial evaluation of the effectiveness for the measures to improve resilience, it is confirmed that core damage frequency can be reduced by applying fracture control concept. The branch probability of the measures to improve resilience proposed in this study is tentatively assigned based on the assumption. This value is expected to be quantified by the forthcoming analyses of the integrity for the reactor vessel structure at ultra-high temperature. The technology developed in this study will be applied for the evaluation of improving resilience of the next generation sodium-cooled fast reactor.

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

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.

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