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論文

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.

論文

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:1 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.

論文

Effect of quenching on molten core-concrete interaction product

北垣 徹; 池内 宏知; 矢野 公彦; Brissonneau, L.*; Tormos, B.*; Domenger, R.*; Roger, J.*; 鷲谷 忠博

Journal of Nuclear Science and Technology, 56(9-10), p.902 - 914, 2019/09

 被引用回数:2 パーセンタイル:34.3(Nuclear Science & Technology)

Characterization of fuel debris is required to develop fuel debris removal tools. Especially, knowledge pertaining to the characteristics of molten core-concrete interaction (MCCI) product is needed because of the limited information available at present. The samples of a large-scale MCCI test performed under quenching conditions, VULCANO VW-U1, by CEA were analyzed to evaluate the characteristics of the surface of MCCI product generated just below the cooling water. As a result, the microstructure of the samples were found to be similar despite the different locations of the test sections. The Vickers hardness of each of the phases in these samples was higher than that of previously analyzed samples in another VULCANO test campaign, VBS-U4. From the comparison between analytical results of VULCANO MCCI test product, MCCI product generated under quenching condition is homogeneous and its hardness could be higher than that of the bulk MCCI product.

論文

First-principles calculation of mechanical properties of simulated debris Zr$$_x$$U$$_{1-x}$$O$$_2$$

板倉 充洋; 中村 博樹; 北垣 徹; 星野 貴紀; 町田 昌彦

Journal of Nuclear Science and Technology, 56(9-10), p.915 - 921, 2019/09

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

福島第一原子力発電所の炉内にある燃料デブリの機械的特性を明らかにするため、模擬デブリである二酸化ウラン・二酸化ジルコニウム溶融物についてその弾性定数と破壊靭性を第一原理計算で評価し、二酸化ジルコニウムの割合の影響を調べ、模擬デブリを用いた実験結果と比較を行った。その結果、実験で観測されている、二酸化ジルコニウム割合の増加に伴う急激な破壊靭性の増加は、二酸化ジルコニウムの複数の相の混在によるものと考えられるという結論を得た。

論文

Knowledge obtained from dismantling of large-scale MCCI experiment products for decommissioning of Fukushima Daiichi Nuclear Power Station

仲吉 彬; 池内 宏知; 北垣 徹; 鷲谷 忠博; Bouyer, V.*; Journeau, C.*; Piluso, P.*; Excoffier, E.*; David, C.*; Testud, V.*

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

A large-scale Molten Core-Concrete Interaction (MCCI) test (VF-U1) under the Fukushima Daiichi Nuclear Power Station (1F) conditions (core material composition, concrete, and decay heat) was conducted at the large MCCI test facility (VULCANO) owned by French Alternative Energies and Atomic Energy Commission (CEA) in France. About 50 kg of simulated debris was melted and brought into contact with concrete to erode concrete under 1F conditions. After cooling, the concrete test section (concrete and MCCI product) was dismantled. Main observations of the structure of solidified pool (crust, porosity, oxide/metal layer, etc.) and of the ablation are given. The technical results obtained herein are summarized, and they provide interesting knowledge that will help with the decommissioning of 1F.

論文

A Laboratory investigation of microbial degradation of simulant fuel debris by oxidizing microorganisms

Liu, J.; 土津田 雄馬; 北垣 徹; 香西 直文; 山路 恵子*; 大貫 敏彦

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 2 Pages, 2019/05

福島第一原子力発電所(FDNPP)の廃止措置を行う上で、燃料デブリの現状を把握することは非常に重要な課題の一つである。事故直後に行われた炉内への海水注入および現在まで続く地下水の流入により、FDNPP周辺環境中の微生物の炉内への侵入が予想される。また、今後予定されている燃料デブリの取り出し作業において、人や機材などの出入りに伴う大気の流入により、新たな微生物の侵入も考えられる。これらの微生物の代謝活動は、炉内構造材の腐食や燃料デブリの分解などの原因となる可能性があり、燃料デブリの現状を正しく把握する上で考慮する必要がある。そこで、本研究では模擬燃料デブリを液体培地中で細菌と混合培養することにより、微生物の代謝による燃料デブリへの影響を観察した。

論文

Large scale VULCANO molten core concrete interaction test considering Fukushima Daiichi condition

Bouyer, V.*; Journeau, C.*; Haquet, J. F.*; Piluso, P.*; 仲吉 彬; 池内 宏知; 鷲谷 忠博; 北垣 徹

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 13 Pages, 2019/03

Fuel debris removal is one of the most important processes for decommissioning a severely damaged nuclear power plant (NPP) such as Fukushima Daiichi NPP (1F). In order to develop relevant removal tools, characteristics of fuel debris are required. In the frame of a JAEA-CEA cooperation, a large-scale MCCI test was performed at the CEA/VULCANO facility using a prototypic metal and oxide corium representative from Fukushima Daiichi unit 1 conditions. Conclusions arising from the material analysis of the selected samples will be relevant for future dismantling operations. This paper deals with the experimental device and process, objective and initial conditions of this MCCI test, and ablation of the concrete quantified in term of volume, depths and velocities. The test section concrete, made with Japanese components, is siliceous with basaltic origin. The main objective of the test was to get a significant ablation leading to an ablation volume ratio of 1.6 in order to produce fuel debris with a composition corresponding to expected conditions in the damaged plant. On a phenomenological point of view, it must be noted that the concrete ablation was clearly anisotropic with a predominantly downwards ablation contrary to previous experiments with silica and limestone concrete.

論文

R&D strategy on mid- and long-term behavior of fuel debris

矢野 公彦; 北垣 徹; 鷲谷 忠博; 宮本 泰明; 小川 徹

Progress in Nuclear Science and Technology (Internet), 5, p.225 - 228, 2018/11

福島第一原子力発電所の廃炉に向けたロードマップによると、燃料デブリ取り出しは2021年、すなわち燃料デブリ生成の10年後に開始される予定である。また燃料デブリは燃料取り出しの終了まで炉内に存在することになる。加えて、炉から取り出された燃料デブリに対して保管が必要になることは想像に難くない。このような事故後の燃料デブリに対する作業を検討するうえで、数十年間の燃料デブリの状態や特性を議論することは不可欠である、そこで原子力機構は燃料デブリの中長期的挙動に関する研究開発戦略を暫定するとともに、この課題に対して国内の大学や他の研究機関と協力し基礎研究を立ち上げている。

論文

Characterization of the VULCANO test products for fuel debris removal from the Fukushima Daiichi Nuclear Power Plant

北垣 徹; 池内 宏知; 矢野 公彦; 荻野 英樹; Haquet, J.-F.*; Brissonneau, L.*; Tormos, B.*; Piluso, P.*; 鷲谷 忠博

Progress in Nuclear Science and Technology (Internet), 5, p.217 - 220, 2018/11

Characterization of the fuel debris is required to develop fuel debris removal tools for the decommissioning of Fukushima Daiichi Nuclear Power Plant (1F). In this study, the VULCANO MCCI test, VBS-U4, was selected as 1F similar conditions and the characteristics of the samples were examined. In the molten pool sample, the round-edged corium-rich oxides region, with diameters of 1-10 mm, is surrounded by a concrete-rich oxide region. It shows convection of the molten pool. Other samples also show the features of the MCCI progression. The main chemical forms of the samples are SiO$$_{2}$$, (U,Zr)O$$_{2}$$, Fe and so on. The microstructure of the samples is heterogeneous structure composed of these phases. The difference in Vickers hardness between the metallic phases and the oxide phases is a distinctive characteristic. It can be noted that the heterogeneous distribution of metallic phases in 1F MCCI products interrupt with the removal operation such as by damaging the core-boring bit.

論文

Mechanical properties of cubic (U,Zr)O$$_{2}$$

北垣 徹; 星野 貴紀; 矢野 公彦; 岡村 信生; 小原 宏*; 深澤 哲生*; 小泉 健治

Journal of Nuclear Engineering and Radiation Science, 4(3), p.031011_1 - 031011_7, 2018/07

Evaluation of fuel debris properties is required to develop fuel debris removal tools for the decommissioning of Fukushima Daiichi Nuclear Power Plant (1F). In this research, the mechanical properties of cubic (U,Zr)O$$_{2}$$ samples containing 10-65% ZrO$$_{2}$$ are evaluated. In case of the (U,Zr)O$$_{2}$$ samples containing less than 50% ZrO$$_{2}$$, Vickers hardness and fracture toughness increased, and the elastic modulus decreased slightly with increasing ZrO$$_{2}$$ content. Moreover, all of those values of the (U,Zr)O$$_{2}$$ samples containing 65% ZrO$$_{2}$$ increased slightly compared to (U,Zr)O$$_{2}$$ samples containing 55% ZrO$$_{2}$$. However, higher Zr content (exceeding 50%) has little effect on the mechanical properties. This result indicates that the wear of core-boring bits in the 1F drilling operation will accelerate slightly compared to that in the TMI-2 drilling operation.

論文

Thermodynamic evaluation of the solidification phase of molten core-concrete under estimated Fukushima Daiichi Nuclear Power Plant accident conditions

北垣 徹; 矢野 公彦; 荻野 英樹; 鷲谷 忠博

Journal of Nuclear Materials, 486, p.206 - 215, 2017/04

AA2016-0278.pdf:0.74MB

 被引用回数:10 パーセンタイル:10.91(Materials Science, Multidisciplinary)

The solidification phases of molten core-concrete under the estimated molten core-concrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core-concrete forms (U,Zr)O$$_{2}$$ and (Zr,U)SiO$$_{4}$$, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO$$_{4}$$ requires a relatively long time. Therefore, the formation of (Zr,U)SiO$$_{4}$$ is limited under quenching conditions. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

論文

Mechanical properties of fuel debris for defueling toward decommissioning

星野 貴紀; 北垣 徹; 矢野 公彦; 岡村 信生; 小原 浩史*; 深澤 哲生*; 小泉 健治

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In the decommissioning of Fukushima Daiichi Nuclear Power Plant (1F), safe and steady defueling work is requested. Before the defueling in 1F, it is necessary to evaluate fuel debris for properties related to the defueling procedure and technology. It is speculated that uranium and zirconium oxide solid solution is one of the major materials of fuel debris in 1F, according to TMI-2 accident experience and the results of past severe accident studies. In this report, mechanical properties of uranium and zirconium oxide solid solution evaluated in the ZrO$$_{2}$$ content range from 10% to 65%.

論文

Direction on characterization of fuel debris for defueling process in Fukushima Daiichi Nuclear Power Station

矢野 公彦; 北垣 徹; 池内 宏知; 涌井 遼平; 樋口 英俊; 鍛治 直也; 小泉 健治; 鷲谷 忠博

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1554 - 1559, 2013/09

For the decommissioning of Fukushima Daiichi Nuclear Power Station (1F), defueling work for the fuel debris in the reactor core of Unit 1-3 is planned to be started within 10 years. Preferential items in the characterization of the fuel debris were identified for the defueling work at 1F, in which the procedure and handling tools were assumed from information of 1F and experience of Three Mile Island Unit 2 (TMI-2) accident. The candidates of defueling tools for 1F were selected from TMI-2 defueling tools. It was found out that they were categorized as 6 groups by their working principles. Important properties on the fuel debris for the defueling were picked up from considering influence of objective materials on their performance. The selected properties are shape, size, density, thermal conductivity, heat capacity, melting point, hardness, elastic modulus, and fracture toughness. In these properties, mechanical properties, i.e. hardness, elastic modulus, fracture toughness were identified as preferential items, because there are few data on that of fuel debris in the past severe accident studies.

論文

Fission product separation from seawater by electrocoagulation method

北垣 徹; 星野 貴紀; 三本松 勇二; 矢野 公彦; 竹内 正行; 五十嵐 武士*; 鈴木 達也*

Journal of Radioanalytical and Nuclear Chemistry, 296(2), p.975 - 979, 2013/05

 被引用回数:4 パーセンタイル:60.1(Chemistry, Analytical)

At the Fukushima Daiichi NPPs, a large amount of seawater containing high activity fission product was accumulated and its treatment has been serious problem. Electrocoagulation method is expected to be part of a useful separation system that can reduce the amount of waste and decrease processing time. In this study, powdered adsorbents, such as ferrocyanide and zeolite, added to seawater containing simulated fission products, and electrocoagulation effect were investigated. As a result, more than 99% of Cs and I were removed. Moreover, rapid solution reactivity with heat was not observed, so the thermal risk of aqueous processing of the aggregation would be low. In addition, thermal analyses showed that the electrocoagulation process did not lead to thermal decomposition. Therefore, in the case electrocoagulation method is applied to decontamination system, it has the potential to thermally stabilize and reduce waste.

論文

Applicability of single mode fiber laser for wrapper tube cutting

涌井 遼平; 北垣 徹; 樋口 英俊; 竹内 正行; 小泉 健治; 鷲谷 忠博

Proceedings of 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference (ICONE-20 & POWER 2012) (DVD-ROM), 7 Pages, 2012/07

Japan Atomic Energy Agency (JAEA) has been developing a fuel disassembly system with reliability for FBR fuel reprocessing. Laser technology has a high cutting performance and stable operation. However, it was hard to apply to the fuel disassembly system in our previous study, because of pin damage and dross adhesion between a wrapper tube and fuel pins. The advance of the laser cutting technology has recently attracted. Development of single mode fiber laser (SMFL) with small diameter of a beam spot has been especially reported. Then, we believed that it has become possible to prevent the dross adhesion in the disassembly. The main purpose of this study is to reevaluate an applicability of laser for the wrapper tube cutting by the basic cutting tests. Concretely, we researched whether cutting conditions such as SMFL etc, have the effects on dross adhesion and pin damage or not and tried to prevent these original matters. As the result, it was demonstrated that the kerf width of SMFL is still thinner than that of multi mode fiber laser (MMFL). The phenomenon is very important to decrease the amount of dross. Therefore, we confirmed that SMFL is suitable for prevention of the original matters and this experimental results shown the new feasibility method of wrapper tube cutting.

論文

Design and fabrication of the FBR fuel disassembly system

北垣 徹; 田坂 應幸; 樋口 英俊; 小泉 健治; 平野 弘康; 鷲谷 忠博; 小林 嗣幸*

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 5 Pages, 2009/06

文部科学省公募研究として、平成18年度から4年計画で実施中の「FBR燃料集合体を対象とした解体・せん断技術開発」において、平成18年度に実施した基本設計に基づき解体システム試験装置を製作した。本報ではこの解体システム試験装置の概略と、模擬燃料集合体を用いた動作試験の概要について報告する。

口頭

燃料集合体解体及び燃料ピンせん断技術の開発,7; 解体システム試験装置の概略

北垣 徹; 鷲谷 忠博; 田坂 應幸

no journal, , 

文部科学省公募研究として、平成18年度から4年計画で実施中の「FBR燃料集合体を対象とした解体・せん断技術開発」において、H18年度に実施した基本設計に基づき解体システム試験装置を製作した。本報ではこの解体システム試験装置の概略と、模擬実用燃料集合体を用いた動作試験の概要について報告する。

口頭

Development of disassembly technology for FBR fuel assembly

北垣 徹; 田坂 應幸; 鷲谷 忠博; 小林 嗣幸

no journal, , 

FaCTプロジェクトの一環として実施中のFBR燃料集合体を対象とした解体技術開発において、機械式解体システム試験装置の製作,一連の解体動作確認を行った。本報ではこれらの成果及び今後の計画について報告する。

口頭

燃料集合体解体及び燃料ピンせん断技術の開発,11; 燃料ピン束移送システムの開発

樋口 英俊; 小泉 健治; 平野 弘康; 北垣 徹; 鷲谷 忠博; 小林 嗣幸*; 田坂 應幸*

no journal, , 

平成19年度に製作した燃料ピンの移送システムに昇降機能の改良を加え、燃料ピンの脱落を防止できる信頼性の高い移送システムを開発した。昇降機能を応用して燃料ピン束を整列させて移送する方法、並びに将来的なシステムとして解体装置から直接、せん断マガジンへ装荷する方式の要素試験結果とその可能性について報告する。

口頭

燃料集合体解体及び燃料ピンせん断技術の開発,10; 解体システム試験装置における基本性能評価試験

北垣 徹; 樋口 英俊; 小泉 健治; 平野 弘康; 田坂 應幸*; 鷲谷 忠博; 小林 嗣幸*

no journal, , 

原子力機構と日本原子力発電は、FaCTプロジェクトの一環としてFBR燃料集合体を対象とした解体技術の開発を行っている。平成20年度には解体システム試験装置の基本性能評価試験を実施し、一連の解体工程の成立性の見通しを得た。本報ではこれらの成果及び今後の計画について報告する。

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