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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.


Universal methodology for statistical error and convergence of correlated Monte Carlo tallies

植木 太郎

Nuclear Science and Engineering, 193, p.776 - 789, 2019/07



平成30年度研究開発・評価報告書; 研究開発課題「原子力安全規制行政への技術的支援及びそのための安全研究」(中間評価)


JAEA-Evaluation 2019-001, 138 Pages, 2019/06




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



Plastic collapse stresses based on flaw combination rules for pipes containing two circumferential similar flaws

長谷川 邦夫; Li, Y.; Kim, Y.-J.*; Lacroix, V.*; Strnadel, B.*

Journal of Pressure Vessel Technology, 141(3), p.031201_1 - 031201_5, 2019/06




吉田 一雄; 玉置 等史; 吉田 尚生; 吉田 涼一朗; 天野 祐希; 阿部 仁

日本原子力学会和文論文誌, 18(2), p.69 - 80, 2019/06



Experimental investigation of decontamination factor dependence on aerosol concentration in pool scrubbing

孫 昊旻; 柴本 泰照; 岡垣 百合亜; 与能本 泰介

Science and Technology of Nuclear Installations, 2019, p.1743982_1 - 1743982_15, 2019/06

Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered. DF dependence in the pool scrubbing with 2.4 m water submergence was investigated by light scattering aerosol spectrometers. It was observed that DF increased monotonically as decreasing particle number concentration in a constant thermohydraulic condition. Two validation experiments were conducted to confirm the observed DF dependence. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found the DF dependence became more significant in higher water submergence.


Development of a function calculating internal dose coefficients based on ICRP 2007 Recommendations

真辺 健太郎; 佐藤 薫; 高橋 史明

BIO Web of Conferences (Internet), 14, p.03011_1 - 03011_2, 2019/05



The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(5), p.432 - 439, 2019/05

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

Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided samples was found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm.


Influence evaluation of sampling methods of the non-destructive examination on failure probability of piping based on probabilistic fracture mechanics analyses

真野 晃宏; 勝山 仁哉; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05



Effect of experimental setting and surface roughness on oxidation behavior of Zry-4 in steam at 1273 K

Negyesi, M.; 天谷 政樹

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

This work deals with oxidation behavior of Zry-4 fuel cladding exposed to steam at 1273 K. The condition corresponds to LOCA. The effect of the specimen surface roughness and experimental setting on the oxidation behavior was investigated by employing two experimental techniques for oxidation tests and metallographic analysis along with hydrogen pick-up measurement. Slower heating rate under steam flow led to significantly slower oxidation rate during the subsequent isothermal exposure. As a consequence, the breakaway was delayed substantially. The effect of the specimen surface roughness on the oxidation behavior seemed to be rather minor under the investigated conditions. On the other hand, hydrogen uptake was found to be substantially affected by both the specimen surface roughness and the tested experimental setting.


Simulation analysis on local damage to reinforced concrete panels subjected to oblique impact by different projectiles, 1; Comparison of impact behavior for rigid projectiles with flat and hemispherical nose shape

Kang, Z.; 永井 穣*; 西田 明美; 坪田 張二; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05



Verification of a probabilistic fracture mechanics code PASCAL4 for reactor pressure vessels

Lu, K.; 勝山 仁哉; Li, Y.; 宮本 裕平*; 廣田 貴俊*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05

Probabilistic fracture mechanics (PFM) is considered as a promising methodology in the integrity assessment of structural components in a nuclear power plant since it can rationally represent the influence parameters in their inherent probabilistic distributions without over-conservativeness. In Japan, a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) has been developed by Japan Atomic Energy Agency, which can be used for structural integrity assessments of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Up till now, many efforts have been made on verifying the PASCAL4 code. Among them, a Japanese working group which is consisted of seven participants from industries, universities and institutes was established to conduct the verification studies. Based on verification activities during the past two years, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs were confirmed with great confidence. This paper summarizes the verification activities in this working group including the verification plan, analysis conditions and results.


Simulation analysis on local damage to reinforced concrete panels subjected to oblique impact by different projectiles, 2; Comparison of impact behavior for soft projectiles with flat and hemispherical nose shape

永井 穣*; Kang, Z.; 西田 明美; 坪田 張二; Li, Y.

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05



Preparedness and response for nuclear or radiological emergency as a designated public corporation

奥野 浩; 岡本 明子; 海老根 典也; 早川 剛; 田中 忠夫

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 15 Pages, 2019/05




森 愛理; 石崎 梓; 普天間 章; 田辺 務; 和田 孝雄; 加藤 貢; 宗像 雅広

保健物理, 54(1), p.45 - 54, 2019/04

Large quantities of radionuclides were released as a result of Fukushima Daiichi Nuclear Power Station accident. It is known that these radionuclides contaminated inside houses as well as outdoor environment. Considering the radiation protection of residents after a nuclear power station accident, it is important to know the influence of radionuclides inside houses to radiation dose to residents. In this study, we investigated removal factors and fractions of fixed contamination of various materials inside houses in Okuma Town, Futaba Town, and Namie Town to assess the contamination level inside house appropriately. Nine kinds of materials, fibers, woods (smooth), woods rough), glasses, concretes (smooth), concretes (rough), plastics, PVCs and metals, were used in examinations. The lowest and the highest removal factors were 23% - 16% of woods (rough) and 79% - 7.7% of glasses, respectively. Removal factors of all materials were higher than 10% which is recommended by Japanese Industrial Standard. The negative correlation was found between removal factors and fractions of fixed contamination. Using this correlation, the decontamination factor, which means the ratio of the activity removed from the surface by one smear sample to the activity of the total surface activity, was proposed. The air dose rate from the contamination inside house was calculated using obtained decontamination factors and removal factor of 10%. In the case using the removal factor of 10%, the air dose rate derived by indoor contamination was approximately 2 times higher than the case using obtained decontamination factors. We found that the air dose rate derived by indoor contamination was much lower than the air dose rate outside house, and the influence of indoor contamination on the external exposure was small.


Plastic collapse stresses for pipes with inner and outer circumferential cracks

Mares, V.*; 長谷川 邦夫; Li, Y.; Lacroix, V.*

Journal of Pressure Vessel Technology, 141(2), p.021203_1 - 021203_6, 2019/04

 パーセンタイル:100(Engineering, Mechanical)

周方向に内外表面亀裂を有する管の塑性崩壊応力は、米国機械学会のボイラーと圧力容器の規格のSection XIのAppendix Cで推定式が記載されている。このAppendix Cの推定式は欠陥形状が同じであれば内外表面亀裂の塑性崩壊応力は同じである。われわれは、管の平均半径を欠陥面と欠陥以外の面の2つの平均半径を考慮し、内外表面亀裂を有する管の塑性崩壊応力を導いた。その結果、外表面欠陥の塑性崩壊応力は、管の厚さが大きく亀裂が深くて長いとき、Appendix Cの推定式は大きく、非安全側になることが分かった。



千葉 悠介; 西山 裕; 椿 裕彦; 岩井 正樹

JAEA-Technology 2019-002, 29 Pages, 2019/03


原子力災害対策特別措置法及び同法「計画等命令」の改正が、2017年10月30日に実施された。この改正への対応のため、楢葉遠隔技術開発センター遠隔機材整備運用課は、原子力機構内原子力緊急事態支援組織として、対象となる機構各施設から選出された要員に対して緊急時対応用遠隔操作資機材の操作訓練を開始した。当該訓練は、偵察用ロボット(クローラベルト使用の走行ロボット・小型)、作業用ロボット(同前・大型、作業機構(腕状又は長尺トング)付)及び小型無線ヘリの3種の機材の操作訓練を一式とし、受講する要員の訓練経験及び熟練度に応じて初級, 中級及び上級の3段階に分けて実施することとした。本報告は、2018年度上期に実施した初級及び中級の訓練のため策定した要員育成プログラムについて述べたものである。



石崎 梓; 普天間 章; 田窪 一也*; 中西 千佳*; 宗像 雅広

JAEA-Data/Code 2018-022, 20 Pages, 2019/03




Failure bending moment of pipes containing multiple circumferential flaws with complex shape

Li, Y.; 東 喜三郎*; 長谷川 邦夫

International Journal of Pressure Vessels and Piping, 171, p.305 - 310, 2019/03

Flaws due to stress corrosion cracking have been detected in piping systems in nuclear power plants. Failure bending moment of a ductile pipe containing a circumferential flaw is predicted using the net-section stress approach according to ASME Code Section XI as a limit load criterion. However, in the current code, the failure bending moment can only be adopted for a pipe containing a single circumferential flaw with constant depth. In this study, a failure estimation method for pipes containing multiple circumferential flaws with complicated shapes was proposed. Furthermore, failure experiments were performed for stainless steel pipes containing two circular circumferential flaws. The failure bending moments obtained from the experiments were compared with the estimated results. Based on the experimental results, it was concluded that the proposed failure estimation method satisfactorily represents the failure behavior of the pipes and can be applied in engineering application.

3027 件中 1件目~20件目を表示