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Improvements on evaluation functions of a probabilistic fracture mechanics analysis code for reactor pressure vessels

Lu, K.; 勝山 仁哉; Li, Y.

Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.


Problems of DPA cross-sections above 20 MeV in FENDL-3.1d found in A-FNS neutronics analysis

権 セロム*; 今野 力; 太田 雅之*; 春日井 敦*

Journal of Nuclear Science and Technology, 57(3-4), p.344 - 351, 2020/03

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



Material balance evaluation of pyroprocessing for minor actinide transmutation nitride fuel

舘野 春香; 佐藤 匠; 津幡 靖宏; 林 博和

Journal of Nuclear Science and Technology, 57(3-4), p.224 - 235, 2020/03

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



New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1; Introduction of boiling water reactor mock-up assembly degradation test programme

Pshenichnikov, A.; 倉田 正輝; Bottomley, D.; 佐藤 一憲; 永江 勇二; 山崎 宰春

Journal of Nuclear Science and Technology, 57(3-4), p.370 - 379, 2020/03

 被引用回数:1 パーセンタイル:34.59(Nuclear Science & Technology)

The new research and development programme of JAEA/CLADS tests complement the previous investigations related to BWR severe accidents. A series of tests aiming at closing the gaps in understanding of the Fukushima Daiichi degradation sequence at each unit. The paper emphasises the problem of control blade degradation, which influences the accident progression at an early stage and shows the approach for thorough investigation of this problem.


Boron chemistry during transportation in the high temperature region of a boiling water reactor under severe accident conditions

三輪 周平; 高瀬 学; 井元 純平; 西岡 俊一郎; 宮原 直哉; 逢坂 正彦

Journal of Nuclear Science and Technology, 57(3-4), p.291 - 300, 2020/03

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



Predictability of a short-term emergency assessment system of the marine environmental radioactivity

川村 英之; 上平 雄基; 小林 卓也

Journal of Nuclear Science and Technology, 57(3-4), p.472 - 485, 2020/03

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



Estimation of uncertainty in lead spallation particle multiplicity and its propagation to a neutron energy spectrum

岩元 大樹; 明午 伸一郎

Journal of Nuclear Science and Technology, 57(3-4), p.276 - 290, 2020/03



Measurements of thermal-neutron capture cross-section of cesium-135 by applying mass spectrometry

中村 詔司; 芝原 雄司*; 木村 敦; 岩本 修; 上原 章寛*; 藤井 俊行*

Journal of Nuclear Science and Technology, 57(3-4), p.388 - 400, 2020/03

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

$$^{135}$$Cs(n,$$gamma$$)$$^{136}$$Cs反応の熱中性子捕獲断面積($$sigma_{0}$$)及び共鳴積分(I$$_{0}$$)を、ガンマ線及びマススペクトロメトリーにより測定した。我々は、$$^{137}$$Cs標準溶液に不純物として含まれている$$^{135}$$Csを利用した。$$^{137}$$Cs溶液中の$$^{135}$$Csを定量するために、$$^{135}$$Csと$$^{137}$$Csの同位対比をマススぺクトロメトリーにより求めた。分析した$$^{137}$$Cs試料を、京都大学複合原子力科学研究所の研究炉の水圧輸送管を用いて中性子照射を行った。照射位置の中性子成分を求めるために、Co/AlとAu/Alモニタも一緒に照射した。$$sigma_{0}$$を求めるために、Gdフィルターを用いて、中性子カットオフエネルギーを0.133eVに設定した。$$^{137}$$Cs, $$^{136}$$Csとモニタの放射能をガンマ線スペクトロメトリーにより測定した。Westcottコンベンションに基づき、$$sigma_{0}$$とI$$_{0}$$を、それぞれ8.57$$pm$$0.25barn及び45.3$$pm$$3.2barnと導出した。今回得られた$$sigma_{0}$$は、過去の測定値8.3$$pm$$0.3barnと誤差の範囲で一致した。


Neutron emission spectrum from gold excited with 16.6 MeV linearly polarized monoenergetic photons

桐原 陽一; 中島 宏; 佐波 俊哉*; 波戸 芳仁*; 糸賀 俊朗*; 宮本 修治*; 武元 亮頼*; 山口 将志*; 浅野 芳裕*

Journal of Nuclear Science and Technology, 57(3-4), p.444 - 456, 2020/03

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



Experimental and analytical investigation of formation and cooling phenomena in high temperature debris bed

堀田 亮年*; 秋葉 美幸*; 森田 彰伸*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk$"o$ld, P.*; et al.

Journal of Nuclear Science and Technology, 57(3-4), p.353 - 369, 2020/03

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

Key phenomena in the cooling states of debris beds under wet cavity conditions were classified into several groups based on the complicated geometry, nonhomogeneous porosity and volumetric heat of debris beds. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. The modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three additional tests, DEFOR-A, PULiMS and REMCOD were employed to validate implemented models. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. It also tries to report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the PULiMS tests.


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(3-4), p.457 - 462, 2020/03

 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。


Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 57(1), p.68 - 78, 2020/01

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

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73 - 85 GWd/t: M-MDA$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10% - 30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520 - 530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.


Measurement of prompt neutron decay constant with spallation neutrons at Kyoto University Critical Assembly using linear combination method

方野 量太; 山中 正朗*; Pyeon, C. H.*

Journal of Nuclear Science and Technology, 57(1-2), p.169 - 176, 2020/01

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



Liquid film behavior and heat-transfer mechanism near the rewetting front in a single rod air-water system

和田 裕貴; Le, T. D.; 佐藤 聡; 柴本 泰照; 与能本 泰介

Journal of Nuclear Science and Technology, 57(1), p.100 - 113, 2020/01

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

The rewetting front propagation may occur when the fuel rod is cooled by the liquid film flow after it is dried out under accident conditions for BWR cores. Our previous study has revealed importance of precursory cooling, defined as a rapid cooling just before the rewetting, which has a significant effect on the propagation velocity. To understand the mechanism of the precursory cooling, we conducted heat transfer experiments using a single heater rod contained inside the transparent glass pipe to measure heat transfer behavior with simultaneous observation using a high-speed camera. The results showed characteristic effects of the wall temperature on the liquid film flow and liquid droplets formation at the rewetting front, i.e. sputtering. Even when the liquid film flows in rivulets under adiabatic condition, horizontally uniformed rewetting front was observed with increasing wall temperature due to enhanced flow resistance by sputtering. This sputtering effect was also confirmed from observations of the liquid film thickness, which increased with approaching the rewetting front. Heat transfer coefficients were predicted roughly well with a single-phase heat transfer correlation with entrance effects, suggesting the thinner thermal boundary layer downstream of the rewetting front may be one of the precursory cooling mechanisms.


Recent activities in the field of reactor physics

福島 昌宏; 東條 匡志*

Journal of Nuclear Science and Technology, 56(12), p.1061 - 1062, 2019/12

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

核分裂による原子炉の根本的な問題を取り扱う炉物理は、革新的な原子炉を含む様々な炉型の安全性や設計研究において重要な役割を果たす。本稿では、炉物理の分野における近年の活動から、Journal of Nuclear Science and Technologyを含む科学誌に発表されたいくつかの優れた研究をまとめる。


Lahar record during the last 2500 years, Chokai Volcano, northeast Japan; Flow behavior, sourced volcanic activity, and hazard implications

南 裕介*; 大場 司*; 林 信太郎*; 國分 陽子; 片岡 香子*

Journal of Volcanology and Geothermal Research, 387, p.106661_1 - 106661_17, 2019/12

 被引用回数:0 パーセンタイル:100(Geosciences, Multidisciplinary)



JENDL/ImPACT-2018; A New nuclear data library for innovative studies on transmutation of long-lived fission products

国枝 賢; 古立 直也; 湊 太志; 岩本 信之; 岩本 修; 中山 梓介; 江幡 修一郎*; 吉田 亨*; 西原 健司; 渡辺 幸信*; et al.

Journal of Nuclear Science and Technology, 56(11-12), p.1073 - 1091, 2019/11

 被引用回数:1 パーセンタイル:29.85(Nuclear Science & Technology)

長寿命核分裂生成核種(LLFP)の核変換技術確立に向けた革新的研究開発に資することを目的とし、新たな核データライブラリJENDL/ImPACT-2018を開発した。開発した核データライブラリは主要なLLFPである$$^{79}$$Se, $$^{93}$$Zr, $$^{107}$$Pd, $$^{135}$$Csおよび周辺核種(計163核種)に対する中性子及び陽子入射の評価済核反応断面積がエネルギー200MeVを上限として格納されている。断面積の評価においては核反応モデルコードCCONEを用いると共に、測定データの乏しい核種やエネルギー領域の断面積を根拠を持って推定するために微視的な核構造理論を積極的に活用した。また、近年RIBF/RIKENにおいて逆運動学を用いて測定された測定データに基づいて主要な核反応モデルパラメータを最適化した。得られたデータは従来手法により求められた既存の核データライブラリJENDL-4.0/HEやTENDL-2017に比べて、安定核種に対する測定データをよく再現することを確認した。


An Experimental investigation of influencing chemical factors on Cs-chemisorption behavior onto stainless steel

西岡 俊一郎; 中島 邦久; 鈴木 恵理子; 逢坂 正彦

Journal of Nuclear Science and Technology, 56(11), p.988 - 995, 2019/11

 被引用回数:1 パーセンタイル:100(Nuclear Science & Technology)



Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

宇田川 豊; 杉山 智之; 天谷 政樹

Journal of Nuclear Science and Technology, 56(11-12), p.1063 - 1072, 2019/11



Development of a laser chipping technique combined with water jet for retrieval of fuel debris at Fukushima Daiichi Nuclear Power Station

山田 知典; 武部 俊彦*; 石塚 一平*; 大道 博行*; 羽成 敏秀; 柴田 卓弥; 大森 信哉*; 黒澤 孝一*; 佐々木 豪*; 中田 正宏*; et al.

Journal of Nuclear Science and Technology, 56(11-12), p.1171 - 1179, 2019/11

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


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