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Case study on sampling techniques using machine learning and simplified physical model for simulation-based dynamic probabilistic risk assessment

久保 光太郎; Zheng, X.; 石川 淳; 杉山 智之; Jang, S.*; 高田 孝*; 山口 彰*

Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 11 Pages, 2020/11



A Comparative study of sampling techniques for dynamic probabilistic risk assessment of nuclear power plants

久保 光太郎; Zheng, X.; 田中 洋一; 玉置 等史; 杉山 智之; Jang, S.*; 高田 孝*; 山口 彰*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10



Degradation prediction of a gamma-ray radiation dosimeter using InGaP solar cells in a primary containment vessel of the Fukushima Daiichi Nuclear Power Station

奥野 泰希; 山口 真史*; 大久保 成彰; 今泉 充*

Journal of Nuclear Science and Technology, 57(4), p.457 - 462, 2020/04

 被引用回数:0 パーセンタイル:100(Nuclear Science & Technology)

優れた高耐放射線性を備えたリン化インジウムガリウム(InGaP)太陽電池は、高放射線量率環境に適用可能な線量計の有力な候補材料になると予想されている。本研究では、InGaP太陽電池を用いた線量計の寿命を予測するために、照射試験及び経験的計算により、InGaP太陽電池の線量信号としての放射線誘起電流に対する少数キャリア拡散長($$L$$)の影響を明らかした。照射試験では、$$gamma$$線線量率の関数としての短絡電流密度($$J_{rm sc}$$)を測定することでInGaP太陽電池の$$L$$を推定した。また、様々な線量率でInGaP太陽電池を検出器として使用した際の動作寿命を、$$L$$と吸収線量の関係に基づく経験式を用いて推定した。この計算結果から、InGaP太陽電池を用いた線量計が福島第一原子力発電所の原子炉格納容器で10時間以上使用可能であり、廃炉に貢献する耐放射線性を有した線量計である可能性が高いことを明らかにした。


Development of semi-implicit particle method for simulating sodium-water chemical reaction

Li, J.*; Jang, S.*; 山口 彰*; 内堀 昭寛; 高田 孝; 大島 宏之

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11



Application of Bayesian optimal experimental design to reduce parameter uncertainty in the fracture boundary of a fuel cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Proceedings of 14th International Conference on Probabilistic Safety Assessment and Management (PSAM-14) (USB Flash Drive), 10 Pages, 2018/09

The reduction of epistemic uncertainty for safety-related events that rarely occur or require high experimental costs is a key concern for researchers worldwide. In this study, we develop a new framework to effectively reduce parameter uncertainty, which is one of the epistemic uncertainties, by using the Bayesian optimal experimental design. In the experimental design, we used a decision theory that minimizes the Bayes generalization loss. For this purpose, we used the functional variance, which is a component of widely applicable information criterion, as a decision criterion for selecting informative data points. Then, we conducted a case study to apply the proposed framework to reduce the parameter uncertainty in the fracture boundary of a non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimen under loss-of-coolant accident (LOCA) conditions. The results of our case study proved that the proposed framework greatly reduced the Bayes generalization loss with minimal sample size compared with the case in which experimental data were randomly obtained. Thus, the proposed framework is useful for effectively reducing the parameter uncertainty of safety-related events that rarely occur or require high experimental costs.


Uncertainty quantification of fracture boundary of pre-hydrided Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Nuclear Engineering and Design, 331, p.147 - 152, 2018/05

 被引用回数:1 パーセンタイル:75.54(Nuclear Science & Technology)

To quantify the fracture boundary uncertainty for non-irradiated, pre-hydrided Zircaloy-4 cladding tube specimens under loss-of-coolant accident conditions at a light-water reactor, data from integral thermal shock tests obtained by an earlier study are analyzed statistically and the fracture boundary is estimated in terms of probability, as follows. First, a method is proposed to obtain the specimens' fracture probability curve as a function of equivalent cladding reacted (ECR) and initial hydrogen concentration using Bayesian inference with a generalized linear model. A log-probit model is used, modified to reflect the effect of the initial hydrogen concentration on the fracture boundary and the ECR evaluation uncertainty, and scaled to improve convergence. Second, using the modified log-probit model, it is shown that the boundary representing a 5% fracture probability with 95% confidence for the pre-hydrided cladding tube sample is higher than 15% ECR, for initial hydrogen concentrations of up to 800 wppm.


Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

成川 隆文; 山口 彰*; Jang, S.*; 天谷 政樹

Journal of Nuclear Materials, 499, p.528 - 538, 2018/02

 被引用回数:3 パーセンタイル:40.19(Materials Science, Multidisciplinary)

For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best model to estimate the fracture probability. It was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.


Development of prototype reactor maintenance, 1; Application to piping system of sodium-cooled reactor prototype

小竹 庄司*; 近澤 佳隆; 高屋 茂; 大高 雅彦; 久保 重信; 荒井 眞伸; 桾木 孝介; 伊藤 隆哉*; 山口 彰*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04



Development of prototype reactor maintenance, 2; Application to piping support of sodium-cooled reactor prototype

荒井 眞伸; 桾木 孝介; 相澤 康介; 近澤 佳隆; 高屋 茂; 久保 重信; 小竹 庄司*; 伊藤 隆哉*; 山口 彰*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04



Numerical study on modeling of liquid film flow under countercurrent flow limitation in volume of fluid method

渡辺 太郎*; 高田 孝; 山口 彰*

Nuclear Engineering and Design, 313, p.447 - 457, 2017/03

 被引用回数:1 パーセンタイル:80.48(Nuclear Science & Technology)



NiO/Nb$$_{2}$$O$$_{5}$$/C hydrazine electrooxidation catalysts for anion exchange membrane fuel cells

坂本 友和*; 増田 晃之*; 吉本 光児*; 岸 浩史*; 山口 進*; 松村 大樹; 田村 和久; 堀 彰宏*; 堀内 洋輔*; Serov, A.*; et al.

Journal of the Electrochemical Society, 164(4), p.F229 - F234, 2017/01

 被引用回数:11 パーセンタイル:36.36(Electrochemistry)

NiO/ Nb$$_{2}$$O$$_{5}$$/C (8:1), (4:1), (2:1), NiO/C, and Ni/C catalysts for hydrazine electrooxidation were synthesized by an evaporation drying method followed by thermal annealing. Prepared catalysts were characterized by X-ray diffraction (XRD), high-angle annular dark field scanning transmission electron microscopy (HAADF-STEM), energy dispersive X-ray spectrometry (EDS), and X-ray absorption fine structure (XAFS). The highest catalytic activity in mentioned above reactionwas found for Ni/C, followed by: NiO/Nb$$_{2}$$O$$_{5}$$/C (8:1), NiO/Nb$$_{2}$$O$$_{5}$$/C (4:1). NiO/Nb$$_{2}$$O$$_{5}$$/C (2:1) whiles NiO/C has almost no activity for hydrazine oxidation. It was explained by oxygen defect of NiO in NiO/ Nb$$_{2}$$O$$_{5}$$/C from XAFS analysis. The selectivity hydrazine electrooxidation as measured by ammonia production resulted in observation that metallic Ni surface facilitates N-N bond breaking of hydrazine, which was confirmed by density functional theory (DFT) calculations.


Study on self-wastage phenomenon at heat transfer tube in steam generator of sodium-cooled fast reactor with consideration of thermal coupling of fluid and structure

小島 早織*; 内堀 昭寛; 高田 孝; 大野 修司; 福田 武司*; 山口 彰*

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 8 Pages, 2016/11



Dynamic and interactive approach to level 2 PRA using continuous Markov process with Monte Carlo Method

Jang, S.*; 山口 彰*; 高田 孝

Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 11 Pages, 2016/10



Mechanism study of hydrazine electrooxidation reaction on nickel oxide surface in alkaline electrolyte by in situ XAFS

坂本 友和*; 岸 浩史*; 山口 進*; 松村 大樹; 田村 和久; 堀 彰宏*; 堀内 洋輔*; Serov, A.*; Artyushkova, K.*; Atanassov, P.*; et al.

Journal of the Electrochemical Society, 163(10), p.H951 - H957, 2016/08

 被引用回数:17 パーセンタイル:29.03(Electrochemistry)

The catalytic process takes place on nickel oxide surface of a Ni oxide nano-particle decorated carbon support (NiO/C). In-situ X-ray absorption fine structure (XAFS) spectroscopy was used to investigate the reaction mechanism for hydrazine electrooxidation on NiO surface. The spectra of X-ray absorption near-edge structure (XANES) of Ni K-edge indicated that adsorption of OH$$^{-}$$ on Ni site during the hydrazine electrooxidation reaction. Density functional theory (DFT) calculations were used to elucidate and suggest the mechanism of the electrooxidation and specifically propose the localization of electron density from OH$$^{-}$$ to 3d orbital of Ni in NiO. It is found that the accessibility of Ni atomic sites in NiO structure is critical for hydrazine electrooxidation. Based on this study, we propose a possible reaction mechanism for selective hydrazine electrooxidation to water and nitrogen taking place on NiO surface as it is applicable to direct hydrazine alkaline membrane fuel cells.


Unusual pressure evolution of the Meissner and Josephson effects in the heavy-fermion superconductor UPt$$_3$$

郷地 順*; 住山 昭彦*; 山口 明*; 本山 岳*; 木村 憲彰*; 山本 悦嗣; 芳賀 芳範; 大貫 惇睦

Physical Review B, 93(17), p.174514_1 - 174514_5, 2016/05

 被引用回数:1 パーセンタイル:91.77(Materials Science, Multidisciplinary)

The Josephson effect between a single crystal UPt$$_3$$ and a conventional superconductor Al has been investigated under pressure. A possible change of the superconducting order parameter has been detected at a critical $$P_c$$ where the symmetry-breaking antiferromagnetic order disappears. The conclusion is also supported by the different behaviors observed in Josephson current and penetration depth.


New AESJ thermal-hydraulics roadmap for LWR safety improvement and development after Fukushima accident

中村 秀夫; 新井 健司*; 及川 弘秀*; 藤井 正*; 梅澤 成光*; 阿部 豊*; 杉本 純*; 越塚 誠一*; 山口 彰*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5353 - 5366, 2015/08

The Atomic Energy Society of Japan developed a New Thermal-Hydraulics Safety Evaluation Fundamental Technology Enhancement Strategy Roadmap (TH-RM) for LWR Safety Improvement and Development after Fukushima-Daiichi Accident through collaboration of utilities, vendors, universities, research institutes and technical support organizations for regulatory body. The revision has been made by three sub working groups (SWGs), by considering the lessons learned from the Fukushima-Daiichi Accident. The safety assessment SWG pursued development of safety assessment computer codes. The fundamental technology SWG pursued safety improvement and risk reduction via improved accident management measures by referring the technical map for severe accident established by severe accident SWG. Twelve important subjects have been identified, and Fact Sheet was developed for each of subjects for research and development. External hazards are also considered how to cope with from thermal-hydraulic safety point of view. This paper summarizes the revised TH-RM with several examples and future perspectives.


Numerical investigation of self-wastage phenomena in steam generator of sodium-cooled fast reactor

Jang, S.*; 高田 孝; 山口 彰*; 内堀 昭寛; 栗原 成計; 大島 宏之

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4275 - 4288, 2015/08



Experimental and numerical reaction analysis on sodium-water chemical reaction field

出口 祥啓*; 高田 孝*; 山口 彰*; 菊地 晋; 大島 宏之

Mechanical Engineering Journal (Internet), 2(1), p.14-00029_1 - 14-00029_11, 2015/02



Numerical study on inert gas behavior in fast reactor primary coolant system; Inert gas accumulation at HPP and consideration of gas elimination system

高田 孝*; 小中 祐至*; 山口 彰*; 伊藤 啓; 大野 修司; 大島 宏之

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11



Numerical quantification of self-wastage phenomena in sodium-cooled fast reactor

Jang, S.*; 高田 孝; 山口 彰*; 内堀 昭寛; 栗原 成計; 大島 宏之

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 8 Pages, 2014/11


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