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論文

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.

論文

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:25.23(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.

論文

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.

論文

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.

論文

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.

論文

ウラン-ジルコニウム-鉄の混合溶融酸化物の凝固時偏析に関する基礎検討

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。

論文

Characterization of the Fukushima Dai-ichi Unit 2 sediments / debris based on the on-site video investigations in comparison to the debris obtained after integral CLADS-MADE-01 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

第24回動力・エネルギー技術シンポジウム講演論文集(USB Flash Drive), 4 Pages, 2019/06

The new data from video investigation of the 1F Unit 2 pedestal debris performed by TEPCO was analysed. The debris features as derived from visual appearance on the video compared with the debris obtained after the CLADS-MADE-01 test. Some speculative conclusions concerning the properties and possible nature of the debris can be made.

論文

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:16.3(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.

論文

Heterogeneity of BWR control blade degradation under steam-starved conditions

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

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

This paper presents recent results on high-temperature control blade degradation at the very beginning phase of a severe accident in BWRs. The large-scale experiment has been performed in JAEA-CLADS laboratory using Large-scale Equipment for Investigation of Severe Accidents in Nuclear reactors (LEISAN). A prototypic sequence of Fukushima Daiichi (1F) Unit 2 has been taken. It has been shown that due to specific conditions happened at Unit 2 the lack of available oxygen allowed metallic melt more flexibility for relocation and inhomogeneous redistribution of melt components due to specially recreated temperature gradient. Phase composition of remained B$$_{4}$$C control blade claddings at different elevations, and phase composition of melt has been investigated by complementary methods and have shown significant difference in elevations together with stratification of metallic components with origination of high melting point layers due to redistribution of steel components and involvement of B and C. It allowed absorber blade residuals with B$$_{4}$$C inside being severely damaged by melting to survive at 1475$$^{circ}$$C and protect B$$_{4}$$C from direct contact with steam.

論文

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.

論文

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

ジルカロイ被覆管とUO$$_{2}$$燃料の高温における溶融過程解析モデルの高度化に資するため、ZrとUO$$_{2}$$の高温反応試験を実施した。UO$$_{2}$$るつぼに、Zr試料を装荷し2173K近傍で加熱を行い、生成した反応相の成長状況や溶融状態、組織変化の観察を行った。試料の中間領域には、上方へ直線状に伸びる相が観測された。この相は、U-Zrの金属溶体相と考えられ、Zr試料中、酸素濃度が少ない方へ選択的に成長したと考えられる。

論文

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

RPV内部の溶融材料の挙動を正確に理解し、SAコードの精度を向上させるために、JUPITERと呼ばれる多相,多物理モデルを備えた新しい計算流体力学(CFD)コードが開発された。それは多相計算のアルゴリズムを最適化した。その上、化学反応もコード内で注意深くモデル化されているので、融解プロセスを正確に扱うことができる。一連の検証と検証の研究が行われており、これらは分析解や以前の実験とよく一致している。JUPITERコードのマルチフィジックスモデルの機能は、関連するシビアアクシデントシナリオにおける溶融材料の挙動を調査するためのもう1つの便利なツールである。

論文

異材溶接継手の界面破断の力学的要因分析

山下 拓哉; 山下 勇人; 永江 勇二

鉄と鋼, 105(1), p.96 - 104, 2019/01

 被引用回数:1 パーセンタイル:66.53(Metallurgy & Metallurgical Engineering)

火力・原子力発電プラントの使用温度条件である550度では、フェライト鋼と溶接材の界面で破断が生じる研究結果が報告されている。本研究では、フェライト鋼への溶接時の入熱量が異なる2種類の異材継手を製作した。溶接には高入熱であるプラズマ溶接および低入熱であるティグ溶接をそれぞれ使用した。改良9Cr-1Mo鋼に形成された熱影響部の組織はプラズマ溶接とティグ溶接とで異なっていた。改良9Cr-1Mo鋼/Alloy 600部を用いて試験片を製作し、550度のクリープ試験を実施した。試験より、ティグ溶接を使用した試験片は界面破断し、プラズマ溶接を使用した試験片は界面破断しない結果が得られた。そのため、熱影響部のひずみ分布計測および有限要素解析を実施し、フェライト鋼に形成される熱影響部の変形挙動に着目し界面破断メカニズムを力学的観点で分析した。各溶接法により製作した異材接手のフェライト鋼の界面近傍に形成する熱影響部の特性の違いにより、界面近傍でのひずみ分布に違いが生じることが分かった。界面破断を回避するためには、フェライト鋼界面近傍にクリープひずみ速度が遅い熱影響部を形成させる必要がある。

論文

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.

論文

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

燃料集合体の酸化及び水素吸収はその後の事故進展挙動に影響を与えることから、PWR燃料集合体では、実効的な水蒸気流量としてg-H$$_{2}$$O/sec/rodという単位が導入されており、事故進展評価の重要なパラメータといて用いられている。一方BWRにおいては、燃料集合体の構成がPWRとは異なることにより、PWRで用いられている規格化された水蒸気流量ではチャンネルボックスの内外での酸化及び水素吸収の差が正確に評価できない。そのため、PWRで用いられているg-H$$_{2}$$O/sec/rodという規格化された水蒸気流量に代わる、適切な評価パラメータがBWRでも必要である。そこで、ジルカロイの水蒸気枯渇条件での酸化及び水素吸収データを取得するため、実機を模擬したBWRバンドル試験体を用いて高温酸化試験を行なった。BWRにおける水蒸気流量を規格化するため、水蒸気流路断面積を考慮したパラメータを検討した。

論文

The CMMR program; BWR core degradation in the CMMR-3 test

山下 拓哉; 佐藤 一憲; 阿部 雄太; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of International Conference on Dismantling Challenges; Industrial Reality, Prospects and Feedback Experience (DEM 2018) (Internet), 11 Pages, 2018/10

2011年に発生した福島第一原子力発電所事故における、燃料集合体の溶融進展挙動については、未だ十分に解明されていない。1979年に発生したスリーマイル島原子力発電所2号炉の事故以降、加圧水型原子炉を中心としたシビアアクシデントについては、炉心溶融の初期挙動や圧力容器破損に関わる個別現象に着目した試験が多数行われてきた。しかし、炉心溶融が進行し、炉心物質が炉心から下部プレナムへと移行する過程に関わる既往研究は少なく、特に、この移行経路に制御棒と複雑な炉心下部支持構造が存在する沸騰水型原子炉(以下、「BWR」という)条件での試験データはほとんどない。本研究では、UO$$_{2}$$ペレットの代りにZrO$$_{2}$$ペレットを用いた燃料集合体規模の試験体に対し、BWR実機で想定される軸方向温度勾配をプラズマ加熱により実現し、高温化炉心のガス透過性および高温化炉心物質の支持構造部への進入と加熱を明らかにするための試験を実施した。その結果、高温化した炉心燃料は、部分的な閉塞を形成するが、残留燃料柱は互いに融着しない傾向が強く、崩壊した場合を含めて気相(及び液相)に対するマクロな透過性を持つことが明らかとなった。

論文

Development of experimental technology for simulated fuel-assembly heating to address core-material-relocation behavior during severe accident

阿部 雄太; 山下 拓哉; 佐藤 一憲; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO$$_{2}$$ with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m $$times$$ 0.3 m $$phi$$) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient ($$>$$ 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.

論文

The CMMR program; BWR core degradation in the CMMR-1 and the CMMR-2 tests

山下 拓哉; 佐藤 一憲; 阿部 雄太; 中桐 俊男; 石見 明洋; 永江 勇二

Proceedings of 12th International Conference of the Croatian Nuclear Society; Nuclear Option for CO$$_{2}$$ Free Energy Generation (USB Flash Drive), p.109_1 - 109_15, 2018/06

2011年に発生した福島第一原子力発電所事故における、燃料集合体の溶融進展挙動については、未だ十分に解明されていない。1979年に発生したスリーマイル島原子力発電所2号炉の事故以降、加圧水型原子炉を中心としたシビアアクシデントについては、炉心溶融の初期挙動や圧力容器破損に関わる個別現象に着目した試験が多数行われてきた。しかし、炉心溶融が進行し、炉心物質が炉心から下部プレナムへと移行する過程に関わる既往研究は少なく、特に、この移行経路に制御棒と複雑な炉心下部支持構造が存在する沸騰水型原子炉(以下、「BWR」という)条件での試験データはほとんどない。本研究では、UO2ペレットの代りにZrO2ペレットを用いた燃料集合体規模の試験体に対し、BWR実機で想定される軸方向温度勾配をプラズマ加熱により実現し、高温化炉心のガス透過性および高温化炉心物質の支持構造部への進入と加熱を明らかにするための試験を実施した。その結果、高温化した炉心燃料は、部分的な閉塞を形成するが、残留燃料柱は互いに融着しない傾向が強く、崩壊した場合を含めて気相(及び液相)に対するマクロな透過性を持つことが明らかとなった。

論文

Application of nontransfer type plasma heating technology for core-material-relocation tests in boiling water reactor severe accident conditions

阿部 雄太; 佐藤 一憲; 中桐 俊男; 石見 明洋; 永江 勇二

Journal of Nuclear Engineering and Radiation Science, 4(2), p.020901_1 - 020901_8, 2018/04

原子力機構では非移行型プラズマ加熱を用いたBWR体系での炉心物質の下部プレナムへの移行挙動(CMR)に着目した試験を検討している。この技術の適用性を確認するために、我々は小規模試験体(107mm$$times$$107mm$$times$$222mm (height))を用いたプラズマ加熱の予備実験を行った。これらの予備実験の結果から、SA(シビアアクシデント)研究への非移行型プラズマ加熱の優れた適用性が確認できた。また我々は、中規模の模擬燃料集合体(燃料ピン50ロッド規模)を準備し、まだ技術的な適用性が確認できていない制御ブレードやCMR事体に関する試験を実施予定である。

論文

Metallurgical investigations on creep rupture mechanisms of dissimilar welded joints between Gr.91 and 304SS

山下 拓哉; 永江 勇二; 菊地 浩一*; 山本 賢二*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04

A dissimilar welded joint was adopted to achieve higher thermal efficiency and economy levels in nuclear and thermal power plants. 2 types of dissimilar welded joints which were different heat input during welding to the ferrite steels were manufactured. The dissimilar welded joints were made of following materials; the modified 9Cr-1Mo steel (Gr. 91) for the ferritic steels, the 304 stainless steel for the austenitic steels and the Inconel 600 for the filler metals, Welding methods for the modified 9Cr-1Mo steel were used Plasma Arc Welding and Gas Tungsten Arc Welding (GTAW), respectively. Creep tests were conducted. Specimens by GTAW failed in base metal part and interface between the modified 9Cr-1Mo steel and Inconel 600. Interface failure mechanisms were analyzed from a perspective of metallurgy which were precipitation and growth of type I carbide and formation of oxide layer.

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