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Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective effect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.


Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06



Mechanical properties database of reactor pressure vessel steels related to fracture toughness evaluation

飛田 徹; 西山 裕孝; 鬼沢 邦雄

JAEA-Data/Code 2018-013, 60 Pages, 2018/11


原子炉圧力容器の健全性を判断する上で、破壊靱性をはじめとする材料の機械的特性は重要な情報となる。本レポートは、日本原子力研究開発機構が取得した中性子照射材を含む原子炉圧力容器鋼材の機械的特性、具体的には引張試験, シャルピー衝撃試験, 落重試験及び破壊靱性試験の公開データをまとめたものである。対象とした材料は、初期プラントから最新プラント相当の不純物含有量及び靱性レベルで製造されたJIS SQV2A(ASTM A533B Class1)相当の5種類の原子炉圧力容器鋼である。また母材に加え、原子炉圧力容器の内張りとして用いられている2種類のステンレスオーバーレイクラッド材の機械的特性データについても記載した。これらの機械的特性データは、材料ごとにグラフで整理するとともに今後のデータの活用しやすさを考慮して表形式でリスト化した。


Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

湯村 尚典; 天谷 政樹

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

 パーセンタイル:100(Nuclear Science & Technology)

To investigate the relationship between the fracture resistance of a cladding tube and the amount of deformation of the cladding tube due to ballooning and rupture during a loss-of-coolant accident (LOCA), four-point-bending tests were performed using non-irradiated Zircaloy-4 cladding tubes which experienced a LOCA-simulated sequence (ballooning, rupture, high temperature oxidation and quench). According to the obtained results, it was found that the maximum bending stress of the cladding tube after the LOCA-simulated sequence, which was defined as the fracture resistance, correlated to the average thickness of prior-$$beta$$ layer in the cladding tube. Based on the average thickness of prior-$$beta$$ layer, the fracture resistance of the cladding tube with ballooning and rupture was expressed as functions of isothermal oxidation time and temperature and the maximum circumferential strain on the cladding tube.


Corrosion behaviour of FeCrAl-ODS steels in nitric acid solutions with several temperatures

高畠 容子; 安倍 弘; 佐野 雄一; 竹内 正行; 小泉 健治; 坂本 寛*; 山下 真一郎

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 9 Pages, 2018/10

事故耐性軽水炉燃料の燃料被覆管として開発されているFeCrAl-ODS鋼の硝酸腐食評価を、使用済燃料再処理工程に対して燃料被覆管腐食生成物が与える影響を評価するために実施した。3mol/L硝酸における腐食試験を、60$$^{circ}$$C, 80$$^{circ}$$C,沸騰条件において実施し、浸漬試験の試験片に対してはXPS分析を行った。沸騰条件にて最も腐食が進展し、腐食速度は0.22mm/yであった。酸化被膜内のFe割合は減少しており、CrとAlの割合は増加していた。腐食試験の結果、FeCrAl-ODS鋼は高い硝酸腐食耐性を持つため、再処理工程中の溶解工程において許容可能であることを確かめた。


OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dost$'a$l, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; 天谷 政樹; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

The benchmark on PCMI was initiated by OECD/NEA Expert Group on Reactor Fuel Performance (EGRFP) in June 2015 and is currently in the latter stages of compiling results and preparing the final report. The aim of the benchmark is to improve understanding and modelling of PCMI amongst NEA member organisations. This is being achieved by comparing PCMI predictions of different fuel performance codes for a number of cases. Two of these cases are hypothetical cases aiming to facilitate understanding of the effects of code-to-code differences in fuel performance models. The two remaining cases are actual irradiations, where code predictions are compared with measured data. During analysis of participants' results of the hypothetical cases, the assumptions for number of radial pellet cracks and the pellet-clad friction coefficient (which can be zero, finite or infinite) were identified to be important factors in explaining differences between predictions once pellet-cladding contact occurs. However, these parameters varied in the models and codes used originally by the participants. This fact led to the extension of the benchmark by inclusion of two additional cases, where the number of radial pellet cracks and three different values of the friction coefficient were prescribed in the case definition. Seven calculations from six organisations contributed results were compared and analysed in this paper.


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.


Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

上羽 智之; 根本 潤一*; 石谷 行生*; 伊藤 昌弘*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

 パーセンタイル:100(Nuclear Science & Technology)



Proceedings of the Research Conference on Cementitious Composites in Decommissioning and Waste Management (RCWM2017); June 20th and 21st, 2017, Tomioka Town Art&Media Center, Tomioka, Futaba, Fukushima, Japan

佐野 雄一; 芦田 敬

JAEA-Review 2017-021, 180 Pages, 2017/11





菅原 隆徳; 辻本 和文

JAEA-Research 2017-011, 35 Pages, 2017/10




Oxidation kinetics of Zry-4 fuel cladding in mixed steam-air atmospheres at temperatures of 1273 - 1473 K

Negyesi, M.; 天谷 政樹

Journal of Nuclear Science and Technology, 54(10), p.1143 - 1155, 2017/10

 被引用回数:2 パーセンタイル:42.02(Nuclear Science & Technology)

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam_air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0 up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation applicable for thermomechanical analysis codes of nuclear power reactor under severe accidents. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transient and post-transient regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transient regime.


Improving the corrosion resistance of silicon carbide for fuel in BWR environments by using a metal coating

石橋 良*; 田邊 重忠*; 近藤 貴夫*; 山下 真一郎; 永瀬 文久

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09



Welding technology R&D of Japanese accident tolerant fuel claddings of FeCrAl-ODS steel for BWRS

木村 晃彦*; 湯澤 翔*; 坂本 寛*; 平井 睦*; 草ヶ谷 和幸*; 山下 真一郎

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09



The Applicability of SiC-SiC fuel cladding to conventional PWR power plant

古本 健一郎*; 渡部 清一*; 山本 晃久*; 手島 英行*; 山下 真一郎; 齋藤 裕明; 白数 訓子

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09



Performance degradation of candidate accident-tolerant cladding under corrosive environment

永瀬 文久; 坂本 寛*; 山下 真一郎

Corrosion Reviews, 35(3), p.129 - 140, 2017/08



Ultrasonic guided wave approach for inspecting concave surface of the laser butt-welded pipe

古澤 彰憲; 西村 昭彦; 武部 俊彦*; 中村 将輝*; 竹仲 佑介*; 西條 慎吾*; 中本 裕之*

E-Journal of Advanced Maintenance (Internet), 9(2), p.44 - 51, 2017/08



Development of core and structural materials for fast reactors

浅山 泰; 大塚 智史

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06



Proceedings of the Research Conference on Post-accident Waste Management Safety (RCWM2016) and the Technical Seminar on Safety Research for Radioactive Waste Storage; November 7th and 8th 2016, LATOV, Iwaki, Fukushima, Japan

本岡 隆文; 山岸 功

JAEA-Review 2017-004, 157 Pages, 2017/03




The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

 被引用回数:3 パーセンタイル:42.29(Nuclear Science & Technology)

In order to investigate the effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 (Zry-4) cladding tube, laboratory-scale experiments on non-irradiated Zry-4 cladding tube specimens were performed under transient-heating conditions which simulate loss-of-coolant-accident (LOCA) conditions by using an external heating method, and the data obtained were compared to those from a previous study where an internal heating method was used. The maximum circumferential strains ($$varepsilon$$s) of the cladding tube specimens were firstly divided by the engineering hoop stress ($$sigma$$). The divided maximum circumferential strains, ${it k}$s, of the previous study, which used the internal heating method, were then corrected based on the azimuthal temperature difference (ATD) in the cladding tube specimen. The ${it k}$s for the external heating method which was used in this study agreed fairly well with the corrected ${it k}$s obtained in the previous study which employed the internal heating method in the burst temperature range below $$sim$$1200 K. Also, the area of rupture opening tended to increase with increasing of the value which is defined as $$varepsilon$$ multiplied by $$sigma$$. From the results obtained in this study, it was suggested that $$varepsilon$$ and the size of rupture opening of a cladding tube under LOCA-simulated conditions can be estimated mainly by using $$sigma$$, $$varepsilon$$ and ATD in the cladding tube specimen, irrespective of heating methods.


Seawater effects on the soundness of spent fuel cladding tube

本岡 隆文; 上野 文義; 山本 正弘

Proceedings of 2016 EFCOG Nuclear & Facility Safety Workshop (Internet), 6 Pages, 2016/09


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