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Solidification and re-melting mechanisms of SUS-B$$_{4}$$C eutectic mixture

墨田 岳大; 北垣 徹; 高野 公秀; 池田 篤史

Journal of Nuclear Materials, 543, p.152527_1 - 152527_15, 2021/01

Fundamental understanding of the high-temperature interaction between stainless steel (SUS) and B$$_{4}$$C is indispensable for estimating and characterizing the fuel debris generated during severe accidents of boiling water reactors (BWR), such as Fukushima Dai-ichi Nuclear Power Station (FDNPS, also referred to as "1F") in Japan. This study aims at systematically characterizing the solidified products of molten SUS-B$$_{4}$$C mixtures by powder X-ray diffraction (PXRD), scanning electron microscopy- energy dispersive X-ray spectroscopy (SEM-EDS), and thermogravimetry-differential thermal analysis (TG-DTA) with a range of the B$$_{4}$$C content relevant to the fuel debris composition expected at 1F, in order to elucidate the solidification and re-melting mechanisms. The results indicated that $$gamma$$-Fe and (Cr,Fe)$$_{2}$$B are the major solidified phases when the B$$_{4}$$C content is below 3 mass%, while (Cr,Fe)$$_{23}$$(C,B)$$_{6}$$ is formed as an additional third phase when the B$$_{4}$$C content exceeds 3 mass%. The solidification of molten SUS-B$$_{4}$$C mixture and re-melting of solidified SUS-B$$_{4}$$C melt are eutectic, which is mainly controlled by the pseudo-binary Fe-B system that is influenced by the C and Cr content and additional minor components such as Mo.


Oxygen potential measurement of (U,Pu,Am)O$$_{2 pm x}$$ and (U,Pu,Am,Np)O$$_{2 pm x}$$

廣岡 瞬; 松本 卓; 加藤 正人; 砂押 剛雄*; 宇野 弘樹*; 山田 忠久*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

(U$$_{0.623}$$Pu$$_{0.350}$$Am$$_{0.027}$$)O$$_{2}$$に対しては1,673, 1,773, 1,873K、(U$$_{0.553}$$Pu$$_{0.285}$$Am$$_{0.015}$$Np$$_{0.147}$$)O$$_{2}$$に対しては1,873, 1,973Kにおいて、酸素ポテンシャルの測定を実施した。測定は、熱天秤と酸素センサーを用いる気相平衡法により実施した。Uの代わりにAmを添加した場合、酸素ポテンシャルは大きく上昇した。同様に、Uの代わりにNpを添加した場合も酸素ポテンシャルは上昇したが、上昇の効果はPuやAmを添加した場合と比べても小さいものであった。酸素ポテンシャルの測定結果について、酸素分圧と定比組成からのずれをプロットし、欠陥化学により解析することで、欠陥反応の種類を推定した。推定した欠陥反応における平衡定数を評価し、AmとNpを平衡定数の中のエントロピーに組み込むことで、酸素ポテンシャルの測定結果を再現する評価式を導出した。


Oxygen self-diffusion in near stoichiometric (U,Pu)O$$_{2}$$ at high temperatures of 1673-1873 K

渡部 雅; 加藤 正人; 砂押 剛雄*

Journal of Nuclear Materials, 542, p.152472_1 - 152472_7, 2020/12

高温における定比組成近傍の(U,Pu)O$$_{2}$$の酸素自己拡散係数を熱重量法と酸素同位体交換法を組み合わせた手法によって測定することに成功した。(U,Pu)O$$_{2}$$の定比組成における酸素拡散の活性化エネルギーを実験データから評価し、その値を248kJ/molと決定した。また、(U,Pu)O$$_{2 pm x}$$の欠陥移動エネルギーを導出し、これらを用いて酸素自己拡散係数を評価した。その結果、実験データと評価結果は良く一致した。


Atomistic modeling of hardening in spinodally-decomposed Fe-Cr binary alloys

鈴土 知明; 高見澤 悠; 西山 裕孝; Caro, A.*; 外山 健*; 永井 康介*

Journal of Nuclear Materials, 540, p.152306_1 - 152306_10, 2020/11



Estimation of reliable displacements-per-atom based on athermal-recombination-corrected model in radiation environments at nuclear fission, fusion, and accelerator facilities

岩元 洋介; 明午 伸一郎; 橋本 慎太郎

Journal of Nuclear Materials, 538, p.152261_1 - 152261_9, 2020/09

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

様々な放射線環境下の材料の照射損傷の指標として、はじき出し原子数(dpa)が用いられている。これまでの照射損傷モデルではNorgertt-Robinson-Torrens(NRT)モデルが標準とされていたが、最新の非熱的再結合補正(athermal-recombination-correction, arc)を考慮したモデルは照射損傷過程をより正確に再現できるため、arcモデルを用いた材料の照射損傷の評価が今後標準となることが期待されている。本研究では、様々な材料に対するarcモデルを粒子・重イオン輸送計算コードPHITSに組み込み、核分裂、核融合及び加速器施設を想定した幅広いエネルギー範囲における再スケーリング係数(NRT-dpa/arc-dpa)を計算した。計算の結果、核施設の照射環境を想定したエネルギー分布を持つ中性子照射、及びエネルギー600MeV-50GeVを持つ陽子照射の場合、各材料の再スケーリング係数は、照射条件によらず、各材料でほぼ同じ値であることがわかった。本成果は、幅広いエネルギー範囲における様々な核施設の従来の放射線損傷評価において、NRT-dpaを再スケーリングするための大変有益な情報となる。


Post-irradiation examinations of annular mixed oxide fuels with average burnup 4 and 5% FIMA

Cappia, F.*; 田中 康介; 加藤 正人; McClellan, K.*; Harp, J.*

Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05

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

We present post-irradiation examination results on two type of annular mixed oxide fuel pins irradiated in the Fast Flux Test Facility (FFTF) sodium cooled reactor to an average burnup between 4% and 5% fission of initial heavy atom (FIMA). The pins differed only from the initial Pu content, which was 22 wt% and 26 wt%, respectively. The overall performance of the pins was excellent, in line with previous historical results. The pins with higher Pu content experienced higher irradiation temperatures which influenced the fission gas release, fuel swelling, and Cs distribution compared to the other pins. All the post-irradiation examinations results are discussed against the irradiation parameters. In particular, the pins with higher initial Pu content, i.e., 26 wt%, experienced higher power that resulted in enhanced fission gas release compared to the other two pins with 22 wt% initial Pu content. For the pins with higher fission gas release, onset of Cs redistribution was observed. The two pins that had lower initial Pu content and burnup showed a Cs axial distribution similar to the as-produced one.


Oxidation kinetics of silicon carbide in steam at temperature range of 1400 to 1800$$^{circ}$$C studied by laser heating

Pham, V. H.; 永江 勇二; 倉田 正輝; Bottomley, D.; 古本 健一郎*

Journal of Nuclear Materials, 529, p.151939_1 - 151939_8, 2020/02

 被引用回数:1 パーセンタイル:100(Materials Science, Multidisciplinary)

As expected for accident tolerant fuels, investigation of steam oxidation for silicon carbide under the conditions beyond design basis accident scenarios is needed. Many studies focused on steam oxidation of SiC at temperatures up 1600$$^{circ}$$C have been conducted and reported in the literature. However, behavior of SiC in steam at temperatures above 1600$$^{circ}$$C still remains unclear. To complete this task, we have designed and manufactured a laser heating facility for steam oxidation at extreme temperatures. With the facility, we report the first results on the steam oxidation behavior of SiC at temperatures range of 1400-1800$$^{circ}$$C for short term exposure of 1-7 h under atmospheric pressure. Based on the mass change of SiC samples, parabolic oxidation rate and linear volatilization rate were calculated. The oxidation layer appears to be maintained at 1800$$^{circ}$$C in steam, but the bubble formation phenomenon suggests other volatilization reactions may limit its life.


Anomalous small-angle X-ray scattering (ASAXS) study of irradiation-induced nanostructure change in Fe-ion beam irradiated oxide dispersion-strengthened (ODS) steel

熊田 高之; 大場 洋次郎; 元川 竜平; 諸岡 聡; 冨永 亜希; 谷田 肇; 菖蒲 敬久; 金野 杏彩; 大和田 謙二*; 大野 直子*; et al.

Journal of Nuclear Materials, 528, p.151890_1 - 151890_7, 2020/01

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



Material characterization of the VULCANO corium concrete interaction test with concrete representative of Fukushima Daiichi Nuclear Plants

Brissonneau, L.*; 池内 宏知; Piluso, P.*; Gousseau, J.*; David, C.*; Testud, V.*; Roger, J.*; Bouyer, V.*; 北垣 徹; 仲吉 彬; et al.

Journal of Nuclear Materials, 528, p.151860_1 - 151860_18, 2020/01

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

In the framework of JAEA-CEA collaboration, experimental studies have been conducted for estimating the material characteristics of corium debris representative of the Fukushima Daiichi nuclear damaged plants. A test has been performed in the VULCANO facility in CEA Cadarache to simulate the concrete corium interaction (CCI) with prototypic corium (using depleted uranium) and concrete of Fukushima Daiichi 1F1 Nuclear Plant. This paper presents the Post Test Analyses on 9 samples representative of the CCI during this test: in the corium pool, in the crusts and at the vertical and horizontal interfaces with the concrete. Analyses have been performed by SEM/EDS, X-Ray Diffraction, complete dissolution and ICP, micro-hardness measurements of the main phases. The major phases encountered are uranium rich and zirconium rich oxides forming nodules from micrometers to millimeters size, chromium-iron rich precipitates of several micrometers, metallic Fe-Ni droplets and chromium-silicon rich filaments in a matrix, likely vitreous, rich in concrete elements: Si, Al, Ca, but containing up to 12 cations. The matrix is the softer oxide phase, when the Cr rich precipitates are the harder. The analyses are consistent with the estimated macroscopic ablation ratio, but do not still explain the important axial ablation observed for this specific basaltic concrete. The different phases formation, distribution and solidification path are discussed. First comparisons are proposed with the former CCI tests with European concretes. These results give helpful insights for the future dismantling of the plant and for a deeper understanding of the CCI process for basaltic concrete.


Stable structure of hydrogen atoms trapped in tungsten divacancy

大澤 一人*; 外山 健*; 波多野 雄治*; 山口 正剛; 渡辺 英雄*

Journal of Nuclear Materials, 527, p.151825_1 - 151825_7, 2019/12

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



The Effect of nitride formation on the oxidation kinetics of Zry-4 fuel cladding under steam-air atmospheres at 1273-1573 K

Negyesi, M.; 天谷 政樹

Journal of Nuclear Materials, 524, p.263 - 277, 2019/10

 被引用回数:1 パーセンタイル:53.3(Materials Science, Multidisciplinary)

The study deals with the oxidation behavior of fuel cladding under mixed steam-air atmospheres. Oxidation tests of Zry-4 were carried out at temperatures of 1273-1573 K. Post-test weight gain measurement along with metallographic examination were conducted to study separately the kinetics of the region where nitrides formed and the nitride-free region. The weight gain coming from the nitride-free region was estimated employing one-dimensional finite difference oxygen diffusion model and measured thicknesses of the metallic part of the oxidized specimen, the columnar oxide and the oxygen stabilized $$alpha$$-Zr(O) as well as the fraction of the columnar oxide at the oxide/metal interface. Consequently, the weight gain related to the nitride formation has been assessed.


Microstructures of ZrC coated kernels for fuel of Pu-burner high temperature gas-cooled reactor in Japan

相原 純; 植田 祥平; 本田 真樹*; 水田 直紀; 後藤 実; 橘 幸男; 岡本 孝司*

Journal of Nuclear Materials, 522, p.32 - 40, 2019/08



First-principles calculation study on phonon thermal conductivity of thorium and plutonium dioxides; Intrinsic anharmonic phonon-phonon and extrinsic grain-boundary-phonon scattering effects

中村 博樹; 町田 昌彦

Journal of Nuclear Materials, 519, p.45 - 51, 2019/06

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



Ultra-high temperature creep rupture and transient burst strength of ODS steel claddings

矢野 康英; 関尾 佳弘; 丹野 敬嗣; 加藤 章一; 井上 利彦; 岡 弘; 大塚 智史; 古川 智弘; 上羽 智之; 皆藤 威二; et al.

Journal of Nuclear Materials, 516, p.347 - 353, 2019/04

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



Primary radiation damage; A Review of current understanding and models

Nordlund, K.*; Zinkle, S. J.*; Sand, A. E.*; Granberg, F.*; Averback, R. S.*; Stoller, R. E.*; 鈴土 知明; Malerba, L.*; Banhart, F.*; Weber, W. J.*; et al.

Journal of Nuclear Materials, 512, p.450 - 479, 2018/12

 被引用回数:56 パーセンタイル:1.92(Materials Science, Multidisciplinary)

あらゆる種類の放射線影響の科学的理解は、一次損傷、すなわち、高エネルギー粒子によって開始された原子弾き出し事象の直後に生成される欠陥から始まる。このレビューでは、過去数十年にわたり一次損傷の性質について繰り返し行われてきた広範な実験およびコンピュータシミュレーションの研究を検討する。我々は、材料における結晶学的または位相的欠陥の生成、ならびに原子混合、すなわち結晶学的位置の原子が他のものと位置を交換することも検討する。我々はまた、このエネルギー粒子の損傷を定量化するための現在の国際標準であるNorgett-Robinson-Torrens(NRT)-dpaの代替案を提供するための最近の取り組みについて考察する。我々は、NRT-dpaを拡張する新しい補完的な変位量推定(athermal recombination corrected dpa: arc-dpa)と原子混合(replacements per atom: rpa)関数を詳細に提示し、それらの利点と限界について議論する。


Measurement of displacement cross sections of aluminum and copper at 5 K by using 200 MeV protons

岩元 洋介; 吉田 誠*; 義家 敏正*; 佐藤 大樹; 八島 浩*; 松田 洋樹; 明午 伸一郎; 嶋 達志*

Journal of Nuclear Materials, 508, p.195 - 202, 2018/09

 被引用回数:5 パーセンタイル:18.88(Materials Science, Multidisciplinary)



Temperature measurement for in-situ crack monitoring under high-frequency loading

直江 崇; Xiong, Z.*; 二川 正敏

Journal of Nuclear Materials, 506, p.12 - 18, 2018/08

 被引用回数:3 パーセンタイル:36.93(Materials Science, Multidisciplinary)



Cavitation damage in double-walled mercury target vessel

直江 崇; 涌井 隆; 木下 秀孝; 粉川 広行; 羽賀 勝洋; 原田 正英; 高田 弘; 二川 正敏

Journal of Nuclear Materials, 506, p.35 - 42, 2018/08

 被引用回数:1 パーセンタイル:73.25(Materials Science, Multidisciplinary)



Recent studies for structural integrity evaluation and defect inspection of J-PARC spallation neutron source target vessel

涌井 隆; 若井 栄一; 直江 崇; 新宅 洋平*; Li, T.*; 村上 一也*; 鹿又 研一*; 粉川 広行; 羽賀 勝洋; 高田 弘; et al.

Journal of Nuclear Materials, 506, p.3 - 11, 2018/08

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



First-principles study of solvent-solute mixed dumbbells in body-centered-cubic tungsten crystals

鈴土 知明; 都留 智仁; 長谷川 晃*

Journal of Nuclear Materials, 505, p.15 - 21, 2018/07

 被引用回数:2 パーセンタイル:52.02(Materials Science, Multidisciplinary)

タングステン(W)は、将来の核融合炉のプラズマ対向材料として有望視されており、W合金の耐放射線性を向上させその機械的特性を改善するための最適な合金成分を選択することが重要な課題である。本研究では、W結晶中の溶媒と溶質の混合ダンベルを第一原理計算によって調査した。その結果、チタン, バナジウム, クロムはW材の延性を低下させる原因となる照射誘起析出を起こさずに、空孔と格子間原子の再結合を促進させるため、照射効果という観点からではW材の合金元素として望ましいことがわかった。

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