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On the degradation progression of a BWR control blade under high-temperature steam-starved conditions

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Mechanical Engineering Journal (Internet), 7(3), p.19-00503_1 - 19-00503_10, 2020/06

High-temperature control blade degradation tests simulating a beginning phase of a severe accident in BWRs has been comprehensively performed in Japan Atomic Energy Agency (JAEA). In the latest test, a mock-up of BWR bundle material has been investigated under postulated Fukushima Dai-Ichi (1F) Unit 2 accident conditions in a complex heating transient scenario including a phase of lack of available steam. The progress in control blade degradation was monitored with help of an in situ video and the detailed analysis of the solidified metallic melt, so-called metallic debris, was carried out by conventional SEM and XRD methods. These results indicated that the composition of the metallic debris at different elevations has been significantly changed due to the redistribution and relocation of steel elements under the influence of B and C, sometimes accompanied by a formation of high-melting-point layers. The results of this paper significantly contribute to the physical understanding of control blade degradation phenomenology during beginning phase of a core degradation for a special case of steam-starved conditions at 1F Unit 2.


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(4), p.370 - 379, 2020/04

 被引用回数:3 パーセンタイル:31.27(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.


Raman characterization of the simulated control blade debris to understand the boric compounds transformations during severe accidents

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Mechanical Engineering Journal (Internet), 7(2), p.19-00477_1 - 19-00477_8, 2020/04

In order to address the challenge of the future Fukushima Dai-Ichi Nuclear Power Station (1F) debris characterization a new Raman spectroscopy investigation of simulated debris obtained after two control blade degradation tests CLADS-MADE-01 and CLADS-MADE-02 has been performed. A mechanism of the B$$_{4}$$C degradation during the beginning phase of a severe accident until approximately 1873 K is described. A sequence of material interactions of B$$_{4}$$C with stainless steel resulted in partial transformation of B$$_{4}$$C granules into pure graphite, that later experienced oxidation with formation of COx gas. Especially this mechanism is active during melting phase in oxidative environment. At the same time boron was associated with formation of new Cr-B-containing solid phases in liquid melt, that continued relocation depleted by Cr and B, which resulted in redistribution of elements within the degrading reactor core. This knowledge would provide new insights for understanding of the absorber blade degradation mechanism under specific accident conditions close to 1F Unit 2 and Unit 3 reactors and especially would be helpful during potential characterization of metallic debris of 1F.


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:0 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.


Oxidation of silicon carbide in steam studied by laser heating

Pham, V. H.; 永江 勇二; 倉田 正輝; 古本 健一郎*; 佐藤 寿樹*; 石橋 良*; 山下 真一郎

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.670 - 674, 2019/09

Silicon carbide (SiC) has recently attracted much attention as a potential material for accident tolerant fuel cladding. To investigate the performance of SiC in severe accident conditions, study of steam oxidation at high temperatures is necessary. However, the study focusing on steam oxidation of SiC at temperatures above 1600$$^{circ}$$C is still certainly limited due to lack of test facilities. With the extreme oxidation/corrosion environment in steam at high temperatures, current refractory materials such as alumina and zirconia would not survive during the tests. Application of laser heating technique could be a great solution for this problem. Using laser heating technique, we can localize the heat and focus them on the test sample only. In this study, we developed a laser heating facility to investigate high-temperature oxidation of SiC in steam at temperature range of 1400-1800$$^{circ}$$C for 1-7 h. The oxidation kinetics is then being studied based on the weight gain and observation on cross-sectioned surface of tested sample using field emission scanning electron microscope. Off-gas measurement of hydrogen (H$$_{2}$$) and carbon monoxide (CO) generated during the test is also being conducted via a sensor gas chromatography. Current results showed that the SiC sample experienced a mass loss process which obeyed paralinear laws. Parabolic oxidation rate constant and linear volatilization rate constant of the process were calculated from the mass change of the samples. The apparent activation energy of the parabolic oxidation process was calculated to be 85 kJ.mol$$^{-1}$$. The data of the study also indicated that the mass change of SiC under the investigated conditions reached to its steady stage where hydrogen generation became stable. Above 1800$$^{circ}$$C, a unique bubble formation on sample surface was recorded.



須藤 彩子; 水迫 文樹*; 星野 国義*; 佐藤 拓未; 永江 勇二; 倉田 正輝

日本原子力学会和文論文誌, 18(3), p.111 - 118, 2019/08

炉心溶融物の凝固過程での冷却速度の違いは燃料デブリ構成成分の偏析に大きく影響する。偏析傾向を把握するため、模擬コリウム(UO$$_{2}$$, ZrO$$_{2}$$, FeO, B$$_{4}$$C, FP酸化物)の溶融/凝固試験を行った。模擬コリウムはAr雰囲気化で2600$$^{circ}$$まで加熱し、2つの冷却速度での降温を行った。(炉冷(平均744$$^{circ}$$C/min)および徐冷(2600$$^{circ}$$C$$sim$$2300$$^{circ}$$C:5$$^{circ}$$C/min、2300$$^{circ}$$C$$sim$$1120$$^{circ}$$C:平均788$$^{circ}$$C/min)元素分析により、炉冷条件および徐冷条件両方の固化後の試料中に3つの異なる組成を持つ酸化物相および1つの金属相が確認された。炉冷条件、徐冷条件ともにこれら3つの酸化物相へのFeO固溶度はおおよそ12$$pm$$5at%であった。この結果はUO$$_{2}$$-ZrO$$_{2}$$-FeO状態図におおよそ一致している。一方、徐冷条件での試料中に、Zrリッチ相の大粒形化が確認された。この相の組成は液相の初期組成と一致しており、遅い凝固中で液滴の連結が起こり、凝集したと評価した。


Features of a control blade degradation observed ${it in situ}$ during severe accidents in boiling water reactors

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

Journal of Nuclear Science and Technology, 56(5), p.440 - 453, 2019/05

 被引用回数:3 パーセンタイル:23.33(Nuclear Science & Technology)

In the present paper new results using ${it in situ}$ video, are presented regarding BWR control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B$$_{4}$$C granules prevented direct steam attack of residual B$$_{4}$$C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.


The Behaviour of materials in case of solidified absorber melt - oxidized BWR channel box interaction revealed after CLADS-MADE-01 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二; 山崎 宰春

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

The paper summarizes the first results of a thorough SEM investigation uncovering the process of channel box wall penetration by Fe-Cr-Ni-B containing melt. The preliminary oxidation of channel box is shown to play an important role on severe accident progression resulted in the suppression of channel box massive destruction. Only one small droplet came out to the other side of channel box. The mechanism of local beginning of oxide layer destruction with subsequent Zircaloy-4 channel box penetration is under discussion.


High-temperature interaction between zirconium and UO$$_2$$

白数 訓子; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05



Validation and verification for the melting and eutectic models in JUPITER code

Chai, P.; 山下 晋; 永江 勇二; 倉田 正輝

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 14 Pages, 2019/03



Steam oxidation of silicon carbide at temperatures above 1600$$^{circ}$$C

Pham, V. H.; 永江 勇二; 倉田 正輝

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 6 Pages, 2018/10

High temperature interaction of chemical vapor deposition SiC with steam was investigated at 1700-1800$$^{circ}$$C for 0.1-3 h in a mixture of steam and argon gas containing 98% of steam at 1 atm. At the investigated conditions, although a dense oxide layer was observed on sample surface, significant mass loss of SiC occurred. Below 1700$$^{circ}$$C, the oxidation kinetics seems to follow the para-linear laws. The apparent activation calculated based on the data of this study is to be 370 kJ/mol. Rapid degradation and bubbling of SiC at 1800$$^{circ}$$C were observed after 1 h oxidation. This suggested that chemical interaction behaviours above 1700$$^{circ}$$C might be changed due to the liquefaction of silica.


High temperature oxidation test of simulated BWR fuel bundle in steam-starved conditions

山崎 宰春; Pshenichnikov, A.; Pham, V. H.; 永江 勇二; 倉田 正輝; 徳島 二之*; 青見 雅樹*; 坂本 寛*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 8 Pages, 2018/10



Thermal conductivity of U-20 wt.%Pu-2 wt.%Am-10 wt.%Zr alloy

西 剛史; 中島 邦久; 高野 公秀; 倉田 正輝; 有田 裕二*

Journal of Nuclear Materials, 464, p.270 - 274, 2015/09

 被引用回数:2 パーセンタイル:73.04(Materials Science, Multidisciplinary)



Corrosion behavior of cold-worked austenitic stainless steels in liquid lead-bismuth eutectic

倉田 有司

Journal of Nuclear Materials, 448(1-3), p.239 - 249, 2014/05

 被引用回数:21 パーセンタイル:7.06(Materials Science, Multidisciplinary)

液体鉛ビスマスと鋼材の両立性は、加速器駆動システムを開発する上で、重要な研究課題である。オーステナイト系ステンレス鋼では、耐照射性向上のため冷間加工を施すことが想定される。本論文では、オーステナイト系ステンレス鋼の液体鉛ビスマス中腐食挙動に及ぼす冷間加工の影響を検討した。溶体化、20%冷間加工, 50%冷間加工した316SS及びJPCA(15Cr-15Ni-Ti)の腐食試験を、酸素濃度をコントロールした液体鉛ビスマス中で実施した。500$$^{circ}$$Cで中間の酸素濃度(1.4$$times$$10$$^{-7}$$wt.%)の鉛ビスマス中では、フェライト化がわずかに認められた。550$$^{circ}$$C, 1000hの低酸素濃度(4.2$$times$$10$$^{-9}$$wt.%)の鉛ビスマス中では、冷間加工度の増加とともに、316SS, JPCAのフェライト化深さは増加した。550$$^{circ}$$Cで高酸素濃度(10$$^{-5}$$wt.%程度)の鉛ビスマスの場合、1000hでは酸化のみが観察されたが、3000hでは酸化,フェライト化とPb-Biの侵入が起こり、冷間加工はフェライト化とPb-Bi侵入を加速した。オーステナイト系ステンレス鋼の冷間加工は鉛ビスマス中の腐食を加速する例が多いことから、耐照射性だけでなく耐食性についても注意することが必要である。


Corrosion behavior of Si-enriched steels for nuclear applications in liquid lead-bismuth

倉田 有司

Journal of Nuclear Materials, 437(1-3), p.401 - 408, 2013/06

 被引用回数:18 パーセンタイル:11.59(Materials Science, Multidisciplinary)

液体鉛ビスマスと鋼材の両立性は、加速器駆動システムや鉛ビスマス冷却炉を開発するうえで、重要な研究課題である。Siを添加した鋼材は、鉛ビスマスとの優れた両立性が期待される。原子力システムへの適用を目的として、2.5wt.%のSiを添加した316SS及び1.5wt.%のSiを添加したMod.9Cr-1Mo鋼が製造された。本論文では、これらの鋼材の液体鉛ビスマス中における腐食試験結果を報告する。腐食試験は550$$^{circ}$$Cで酸素濃度を2.5$$times$$10$$^{-5}$$wt.%及び4.4$$times$$10$$^{-8}$$wt.%にコントロールした2条件で実施された。316SSへの2.5wt.%のSi添加及びMod.9Cr-1Mo鋼への1.5wt.%のSi添加は、高酸素濃度の鉛ビスマス中では形成する酸化膜の厚さを減少させる効果があった。これに対して、10$$^{-8}$$wt.%程度の低酸素濃度の鉛ビスマス中では、形成した酸化膜は、Ni溶出あるいはPb, Biの侵入を防ぐ十分な防護性を保持してはいないことがわかった。


Progress review of research and development on accelerator driven system in JAEA

大井川 宏之; 辻本 和文; 佐々 敏信; 倉田 有司; 武井 早憲; 斎藤 滋; 西原 健司; 大林 寛生; 菅原 隆徳; 岩元 大樹

KURRI-KR(CD)-40 (CD-ROM), p.16 - 30, 2013/00



Development of aluminum powder alloy coating for innovative nuclear systems with lead-bismuth

倉田 有司; 佐藤 英友*; 横田 仁志*; 鈴木 徹也*

Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.177 - 188, 2012/12

液体鉛ビスマスは、長寿命放射性核種の核変換処理を目的とした加速器駆動システムや高速炉などの革新的原子力システムにおいて使用することが検討されている。このシステムの課題の一つは、鉛ビスマスによる腐食に対する鋼材の耐食性確保である。本研究では、鋼材の耐食性確保のため、Al粉末合金被覆法の開発を行った。Al, Ti, Fe粉末から作製したシート材をSUS316基材に載せ、レーザー加熱によって、Al合金被覆を施した。被覆層のAl濃度の制御は、シート材の組成,レーザーの走査速度などを調整することにより、可能となった。被覆を施した試験片を供試材として、酸素濃度を制御した550$$^{circ}$$Cの鉛ビスマス中で、1,000h及び3,000hの腐食試験を行った。腐食試験の結果、レーザー走査速度が遅く、Al濃度が5$$sim$$8mass%の被覆が、SUS316で観察された鉛ビスマスによる激しい腐食を防いでいることがわかった。


Development of Al-alloy coating for advanced nuclear systems using lead alloys

倉田 有司; 横田 仁志*; 鈴木 徹也*

Journal of Engineering for Gas Turbines and Power, 134(6), p.062902_1 - 062902_7, 2012/06

 被引用回数:4 パーセンタイル:69.15(Engineering, Mechanical)

鉛合金を用いる原子力システムは、安全性の高い高速炉システム概念の一つとして有望であるが、液体鉛合金は高温で鋼材に対する腐食性が強いため、鋼材の耐食性を改善することが必要である。本論文は、液体鉛ビスマスを用いた原子力システムのために開発したAl, Ti, Fe粉末を用いレーザービーム加熱を利用したAl合金被覆法について記述している。このAl合金被覆において、形成される主な欠陥は、表面欠陥とクラックである。これらの欠陥をなくすためには、レーザービーム走査速度を下げること及び被覆層のAl濃度の調整が有効である。550$$^{circ}$$Cの液体鉛ビスマス中での腐食試験により、316SS上のAl合金被覆層は、被覆なしの316SSで顕著であった粒界腐食や鉛ビスマスの侵入を防いでいることがわかった。Al合金被覆の優れた耐食性は、鉛ビスマス中で再生される薄いAl酸化膜によっていることを明らかにした。Al合金被覆層の健全性と耐食性保持の観点から、適切なAl濃度の範囲は4-12wt%であることを示した。


Development of aluminum-alloy coating on type 316SS for nuclear systems using liquid lead-bismuth

倉田 有司; 横田 仁志*; 鈴木 徹也*

Journal of Nuclear Materials, 424(1-3), p.237 - 246, 2012/05

 被引用回数:11 パーセンタイル:27.12(Materials Science, Multidisciplinary)

加速器駆動核変換システム等の液体鉛ビスマスを使用する原子力システムのために、Al, Ti, Fe粉末とレーザービーム加熱を用いたAl合金被覆法を開発した。ここで開発した被覆処理において形成される主な欠陥は、表面欠陥とクラックであった。これらの欠陥をなくすために被覆条件の最適化が行われた。550$$^{circ}$$Cで10$$^{-6}$$$$sim$$10$$^{-3}$$ wt%に酸素濃度を制御した液体鉛ビスマス中で、3000hの静的腐食試験を実施した。その結果、316ステンレス鋼上のAl合金被覆層は、被覆なしの316ステンレス鋼で認められた激しい腐食(Ni溶出,粒界腐食,鉛ビスマスの侵入等)を防いでいることがわかった。最適条件で被覆した表面欠陥やクラックのないAl合金被覆層は、550$$^{circ}$$Cの液体鉛ビスマス中で優れた耐食性を示す。


Present status for research and development on accelerator driven system in JAEA

辻本 和文; 大井川 宏之; 倉田 有司; 西原 健司; 菅原 隆徳; 武井 早憲; 斎藤 滋; 大林 寛生; 岩元 大樹

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 8 Pages, 2011/12


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