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Numerical simulation method using a Cartesian grid for oxidation of core materials under steam-starved conditions

山下 晋; 佐藤 拓未; 永江 勇二; 倉田 正輝; 吉田 啓之

Journal of Nuclear Science and Technology, 60(9), p.1029 - 1045, 2023/09

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

We newly developed a detailed simulation method for the oxide layer growth/recession under steam-starved conditions using computational fluid dynamics (CFD) methodologies to elaborate the understanding of failure conditions of fuel assemblies during severe accidents. The new method uses the concept of the distance function in a Cartesian grid and is implemented in the original multiphase/multicomponent CFD code named JUPITER (JAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research). A distance calculation of the normal direction from the interface is generally difficult in a Cartesian grid. However, the distance function can give a distance normal to the surface of materials by referring to the value of the function. Thus, the growth/recession calculations, which require the distance normal to the interface, become very easy. We checked the availability of JUPITER, considering these models against the verification and validation problems. As a result, we confirmed that JUPITER gives good results, which may contribute to understanding the progress of core degradation under steam-starved conditions.


Study on chemical interaction between UO$$_{2}$$ and Zr at precisely controlled high temperatures

白数 訓子; 佐藤 拓未; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 60(6), p.697 - 714, 2023/06

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

ジルカロイ被覆管とUO$$_{2}$$燃料の溶融反応のメカニズム解明に資するため、温度誤差が可能な限り最小となるよう検討を行い、1840$$^{circ}$$Cから2000$$^{circ}$$Cの範囲で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、アルゴン雰囲気中加熱を行い、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。1890 $$^{circ}$$Cから1930 $$^{circ}$$Cで加熱した試料は、丸く変形しており、$$alpha$$-Zr(O)相と、少量のU-Zr-O溶体相で形成されていた。1940$$^{circ}$$C以上で加熱した試料は大きく変形し、急激に溶体形成反応が進行する様子が観測された。U-Zr-O溶体相の形成反応はZr(O)中の酸素濃度に依存し、酸素濃度の低いZr(O)へ反応はどんどん進展する。そして酸素含有量が高いZr(O)中では、U-Zr-O溶体相の生成が抑制されることが確認された。


Thermodynamic analysis for solidification path of simulated ex-vessel corium

佐藤 拓未; 永江 勇二; 倉田 正輝; Quaini, A.*; Gu$'e$neau, C.*

CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 79, p.102481_1 - 102481_11, 2022/12

 被引用回数:0 パーセンタイル:0.01(Thermodynamics)

Investigation of the primary containment vessel inside the Fukushima Daiichi Nuclear Power Station showed that a significant amount of the molten corium reached the bottom of the pedestal region. The molten corium and concrete likely caused a complex interaction called Molten Corium Concrete Interaction. The solidification hysteresis of these ex-vessel debris significantly influences its properties. We performed a thermodynamic analysis using the TAF-ID database to infer the solidification path of U-Zr-Al-Ca-Si-O molten corium, which was chosen for a prototypic system of ex-vessel debris. The solidification path for the CaO-rich sim-corium showed that (i) melting as a single liquid phase above 2430 K, (ii) selective solidification of the oxide-rich corium mainly consisted of fuel materials, and (iii) solidification of the remaining materials as a silicate matrix. In contrast, the solidification path for the SiO$$_{2}$$-rich corium indicated that (i) formation of liquid miscibility gap above 2200 K between U-rich and Zr-rich oxidic melts, (ii) individual precipitation of solid phases in each liquid phase.


An Investigation of the microstructure and phase composition of the Zr bearing metallic debris in a bypass channel of a BWR fuel after the exothermic reaction in the CLADS-MADE-04 test

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR2022) (Internet), 4 Pages, 2022/10

CLADS-MADE-04は、下部コア領域での溶融伝播挙動の理解を目的としたシリーズの次のテストである。この寄稿では、電子プローブ マイクロアナライザー(EPMA)によって調査された金属破片の微細構造を含む試験後分析の最近の結果について説明する。テスト中、制御棒ブレードの溶融は、比較的ゆっくりと(数cm/分)急激な強い熱放出の波で発生し、最も高温の領域から、劣化しつつある制御棒ブレードとチャネルボックスに沿って下方に広がり、ジルカロイ-4で作られた壁を消費した。サンプル支持板にも大きな損傷が発生した。このような金属リッチ破片の微細構造の調査により、強化された局所コア劣化のメカニズムを理解できるようになる。EPMAによる相同定を徹底した上で、放熱性が高く周囲への拡散の可能性があることを確認する必要がある。Fe-B共晶デブリとZr-Fe共晶デブリの違いについて概説する。これは、下部炉心プレートのメルトスルーと、下部プレナムへのZr-Fe溶融材料の進行の可能性を理解するために特に重要である。


Step-by-step challenge of debris characterization for the decommissioning of Fukushima-Daiichi Nuclear Power Station (FDNPS)

倉田 正輝; 奥住 直明*; 仲吉 彬; 池内 宏知; 小山 真一

Journal of Nuclear Science and Technology, 59(7), p.807 - 834, 2022/07

 被引用回数:5 パーセンタイル:93.29(Nuclear Science & Technology)

東京電力福島第一原子力発電所(1F)の廃炉に向けて、事故直後から、燃料デブリの特性評価について様々な試みが行われてきた。本レビューでは、それらを概説する。事故直後の数年間は、1F現場(特に損傷した1, 2, 3号機の建屋内部)から得られる知見は極めて限定的であった。燃料デブリのおよその所在はミューオントモグラフィーで調査され、その特性は、加圧水型軽水炉である米国スリーマイル原発事故調査の結果等の過去知見に基づいて概略推定された。その後、各種の内部調査ロボットが開発され、2017年より、原子炉格納容器内部の調査が開始された。その結果、3つの号機とも、当初、加圧水型軽水炉の典型的事故シナリオに基づいて予想されたものと異なる炉心破損状態とデブリ堆積状態であることがわかってきた。また、調査ロボットの付着物から微量のウラン含有粒子が回収され、燃料デブリにつながる情報を得るために、その特性分析が継続している。これらと並行して、OECD/NEAなどにおいて、事故シナリオの検討等について国際協力も進んでいる。今後は、典型事故条件に基づくイメージではなく、現場で得られた知見に基づいて、1F事故シナリオの理解を深め、1F燃料デブリの特性を評価していくことが重要である。


Ten years of Fukushima Dai-ichi post-accident research on the degradation phenomenology of the BWR core components

Pshenichnikov, A.; 柴田 裕樹; 山下 拓哉; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 59(3), p.267 - 291, 2022/03

 被引用回数:1 パーセンタイル:19.33(Nuclear Science & Technology)

The paper reviews the results of the JAEA and some International activities over the last ten years of research on the understanding of the core components melting and debris formation in boiling water reactors.


Raman investigation of the CLADS-MADE-02 test debris to confirm the mechanism of the volatile and non-volatile boron compounds formation

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Proceedings of TopFuel 2021 (Internet), 12 Pages, 2021/10

The results of the several recent tests performed in JAEA/CLADS are outlined in this paper. However, particular point of this work is focused on the interesting effect that was found on the debris, that contained partially reacted B$$_{4}$$C (control blade debris). A Raman investigation of the control blade metallic debris helped to refine the governing mechanism of the B-compounds formation and transport, which is probably specific mostly for BWRs due to unique bundle configuration and materials morphology. All these factors may directly influence the accident progression in BWR and influence the final debris properties.


A BWR control blade degradation observed in situ during a CLADS-MADE-02 test under Fukushima Dai-Ichi Unit 3 postulated conditions

Pshenichnikov, A.; 倉田 正輝; 永江 勇二

Journal of Nuclear Science and Technology, 58(9), p.1025 - 1037, 2021/09

 被引用回数:2 パーセンタイル:36.99(Nuclear Science & Technology)

The paper summarizes the results of the control blade degradation test CLADS-MADE-02 performed in JAEA. The test focused at the beginning phase of the accident at Fukushima Dai-Ichi (1F) Unit 3. The investigation provided important data, especially on the temperature history, exhaust gas measurement and in situ video of metallic debris formation and relocation to the colder elevations under the test scenario, which reproduced oxidizing conditions during the initial phase of the 1F Unit 3 reactor heat-up. Based on the test results, some decommissioning related conclusions concerning the formation of new B-rich phases containing Cr and Fe were made.


Steam oxidation of silicon carbide at high temperatures for the application as accident tolerant fuel cladding, an overview

Pham, V. H.; 倉田 正輝; Steinbrueck, M.*

Thermo (Internet), 1(2), p.151 - 167, 2021/09

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000$$^{circ}$$C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600$$^{circ}$$C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.


「廃炉・汚染水対策事業費補助金(燃料デブリの分析精度の向上及び熱挙動の推定のための技術開発)」に係る補助事業; 2020年度最終報告

小山 真一; 中桐 俊男; 逢坂 正彦; 吉田 啓之; 倉田 正輝; 池内 宏知; 前田 宏治; 佐々木 新治; 大西 貴士; 高野 公秀; et al.

廃炉・汚染水対策事業事務局ホームページ(インターネット), 144 Pages, 2021/08



Features of a BWR neutron absorber melt relocation in an oxidative environment during the CLADS-MADE-02 test

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 7 Pages, 2021/08

The work on the understanding of the accident progression at the Units of the Fukushima Dai-Ichi Nuclear Power Plant (1F) is ongoing. This contribution gives a part of detailed investigations of the control blade melt propagation downwards through the prototypic BWR bundle assembly during the CLADS-MADE-02 test, where the conditions of the 1F Unit 3 was simulated. Interesting features emerged in an oxidative environment.


Comparison of the observed Fukushima Dai-ichi Unit 2 debris with simulated debris from the CLADS-MADE-01 control blade degradation test

Pshenichnikov, A.; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 58(4), p.416 - 425, 2021/04

 被引用回数:9 パーセンタイル:78.91(Nuclear Science & Technology)

The paper describes the attempt of comparison of the simulated test CLADS-MADE-01 debris with the observed in the Unit 2. Similarities between them allowed to make conclusions on their possible source. During the test under postulated 1F Unit 2 simulated conditions a complex behaviour of the test sample with formation of mostly three types of debris was observed. A possible mechanism of stone-like debris formation in 1F case is discussed. The results of this paper broaden our understanding of the metallic debris properties after core degradation for a special case of steam-starved conditions at 1F Unit 2.


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.


Segregation behavior of Fe and Gd in molten corium during solidification progress

須藤 彩子; Meszaros, B.*; Poznyak, I.*; 佐藤 拓未; 永江 勇二; 倉田 正輝

Journal of Nuclear Materials, 533, p.152093_1 - 152093_8, 2020/05

 被引用回数:4 パーセンタイル:48.93(Materials Science, Multidisciplinary)

For a criticality assessment of the fuel debris generated by the Fukushima Daiichi Nuclear Power Plant accident, knowing the segregation of neutron absorber materials, ${it i.e.}$, Gd, B, and Fe, in the fuel debris is necessary. Although B may mostly evaporate during melting, Fe and Gd are expected to remain in the molten corium. To understand the redistribution behavior of Gd and Fe during the solidification of the molten corium, solidification experiments with simulated corium (containing UO$$_{2}$$, ZrO$$_{2}$$, FeO, and Gd$$_{2}$$O$$_{3}$$ with a small amount of simulated fission products such as MoO$$_{3}$$, Nd$$_{2}$$O$$_{3}$$, SrO, and RuO$$_{2}$$) were performed using a cold crucible induction heating method. The simulated corium was slowly cooled from 2,500$$^{circ}$$C and solidified from the bottom to the top of the melt. An elemental analysis analysis of the solidified material showed that the Fe concentration in the inner region increased up to approximately 3.4 times that in the bottom region. This suggested that FeO might be concentrated in the residual melt and that, consequently, the concentration of Fe increased in the later solidification region. On the contrary, the Gd concentration in the periphery region was found to be approximately 2.0 times higher than that in the inner region, suggesting the segregation of Gd in the early solidified phase. No significant segregation was observed for the simulated fission products.


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:10 パーセンタイル:66.42(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:12 パーセンタイル:88.42(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.


Advances in fuel chemistry during a severe accident; Update after Fukushima Daiichi Nuclear Power Station (FDNPS) accident

倉田 正輝; 逢坂 正彦; Jacquemain, D.*; Barrachin, M.*; Haste, T.*

Advances in Nuclear Fuel Chemistry, p.555 - 625, 2020/00



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.


Overview of accident-tolerant fuel R&D program in Japan

山下 真一郎; 井岡 郁夫; 根本 義之; 川西 智弘; 倉田 正輝; 加治 芳行; 深堀 智生; 野澤 貴史*; 佐藤 大樹*; 村上 望*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.206 - 216, 2019/09


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