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論文

Chemical forms of uranium evaluated by thermodynamic calculation associated with distribution of core materials in the damaged reactor pressure vessel

池内 宏知; 矢野 公彦; 鷲谷 忠博

Journal of Nuclear Science and Technology, 57(6), p.704 - 718, 2020/06

 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から取り出された燃料デブリへの効果的な処置方策を提案する上では、燃料デブリ中でUがとりうる化学形についての詳細な調査が不可欠である。特に、アクセス性に乏しい圧力容器内に残留する燃料デブリに関する情報が重要である。本研究では、圧力容器内燃料デブリ中、特にマイナー相におけるUの化学形を評価することを目的とし、1F-2号機の事故進展での材料のリロケーション及び環境変化を考慮した熱力学計算を実施した。組成,温度,酸素量といった計算条件は、既存の事故進展解析の結果から設定した。計算の結果、Uの化学形はFeとOの量によって変化し、Feの少ない領域で$$alpha$$-(Zr,U)(O)、Feの多い領域でFe$$_{2}$$(Zr,U) (Laves相)の生成が顕著であった。還元性条件で生成するこれらの金属相中には数パーセントのUが移行しており、燃料デブリの処置において核物質の化学分離を考慮する場合はこれらの相の生成に留意すべきと考えられる。

論文

Leaching behavior of prototypical Corium samples; A Step to understand the interactions between the fuel debris and water at the Fukushima Daiichi reactors

仲吉 彬; Jegou, C.*; De Windt, L.*; Perrin, S.*; 鷲谷 忠博

Nuclear Engineering and Design, 360, p.110522_1 - 110522_18, 2020/04

 被引用回数:1 パーセンタイル:24.17(Nuclear Science & Technology)

Simulated in-vessel and ex-vessel fuel debris, fabricated in the Colima experimental facility set up in the PLINIUS platform at CEA Cadarache, were selected and leaching experiments were carried out under oxidizing conditions. In parallel, geochemical modeling was performed to better understand the experimental concentrations, pH evolutions and secondary phase's formation. Finally, the Fractional Release Rates of the (U, Zr)O$$_{2}$$ matrix for the two kinds of samples (in-vessel and ex-vessel) were found to be close to or one order of magnitude lower than that of SF under oxidizing conditions (from 10$$^{-6}$$ to 10$$^{-7}$$ per day), but the release processes are different.

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

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

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

Current situation of OECD/NEA, Preparatory Study on Analysis of Fuel debris (PreADES) project

仲吉 彬; Bottomley, D.; 鷲谷 忠博

Proceedings of 56th Annual Meeting on Hot Laboratories and Remote Handling (HOTLAB 2019) (Internet),  ( ), 3 Pages, 2019/00

"Preparatory Study on Analysis of Fuel debris" (PreADES) project was recommended by the Senior Expert Group (SEG) on Safety Research Opportunities post-Fukushima (SAREF) as a "near-term project". The PreADES project will summarise the collected knowledge and expertise of debris characterisation and identify the needs for debris analyses that will most contribute to the decommissioning of 1F. The project also aims to improve the understanding of severe accidents and reactor safety assessments as well as creating appropriate and optimal methodologies for future debris sampling, retrieval, and storage. Consequently, the project provides important input for a future international project of sample examination based on "long-term considerations".

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

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.

報告書

使用済燃料プールから取り出した損傷燃料等の処理方法の検討; 平成25、26年度成果概要報告(受託研究)

飯嶋 静香; 内田 直樹; 田口 克也; 鷲谷 忠博

JAEA-Review 2015-018, 39 Pages, 2015/11

JAEA-Review-2015-018.pdf:3.95MB

福島第一原子力発電所の使用済燃料プールの燃料には、海水、コンクリート等の不純物の付着・同伴に加え、落下したガレキによる損傷の可能性もある。これらの損傷燃料等について、再処理が可能か否かを判断するための指標を整備することを目的として、漏えい燃料等の処理経験を有する東海再処理施設並びに海外再処理施設の処理実績、福島第一原子力発電所のプール燃料の貯蔵状況及び点検・調査結果等を整理した上で、損傷燃料等を再処理する際の技術的課題を摘出するとともに、必要な研究要素を整理した。また、研究要素に関する試験・検討結果に基づき、再処理可否を判断するための分別指標を整備した。

論文

Study of treatment scenarios for fuel debris removed from Fukushima Daiichi NPS

鷲谷 忠博; 矢野 公彦; 鍛治 直也; 山田 誠也*; 紙谷 正仁

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

福島第一原子力発電所の燃料デブリ取出し後の処置については、燃料デブリ取出し開始時までにデブリの処置の選択決定に係る一定の議論が必要になるものと想定し、それまでに各シナリオの比較評価に用いる情報や比較評価の進め方を決める必要がある。そのため、本検討では燃料デブリの取出し後の処置シナリオの検討に向けた技術的要件の整理として、各処置シナリオ案の得失評価を行った。評価の結果、総合すると技術課題は有するものの経済性、廃棄物発生量の面で有利なシナリオは長期保管及び直接処分と推定された。一方、安定化処理、湿式処理、乾式処理は経済性、廃棄物発生量の面で不利と推定された。

論文

Property measurements and inner state estimation of simulated fuel debris

廣岡 瞬; 加藤 正人; 森本 恭一; 鷲谷 忠博

Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 8 Pages, 2014/08

福島第一原子力発電所におけるシビアアクシデント以来、炉心から溶融燃料を取り出すための技術開発が行われているが、炉心へのアクセス、燃料デブリの切断、臨界安全管理、フィッサイルの推定、取り出し後の保管等の課題が残っている。本研究では、模擬燃料デブリを用いた解析から燃料デブリの物性を評価し、温度分布や密度分布のような事故後経時変化するデブリの内部状況の推定を行った。融点,熱伝導率,熱膨張等の物性は、UO$$_2$$及びジルカロイから製作した模擬デブリによって取得した。

論文

Dissolution behavior of (U,Zr)O$$_{2}$$-based simulated fuel debris in nitric acid

池内 宏知; 石原 美穂; 矢野 公彦; 鍛治 直也; 中島 靖雄; 鷲谷 忠博

Journal of Nuclear Science and Technology, 51(7-8), p.996 - 1005, 2014/07

 被引用回数:5 パーセンタイル:51.92(Nuclear Science & Technology)

To explore the possibility of dissolving fuel debris as a potential pre-treatment for waste treatment, dissolution tests of U$$_{1-}$$$$_{x}$$Zr$$_{x}$$O$$_{2}$$ and (U,Pu)$$_{1-}$$$$_{x}$$Zr$$_{x}$$O$$_{2}$$ were carried out in 6 M HNO$$_{3}$$ at 353 K. While the U and Zr indicated congruent leaching from the simulated debris with U-rich compositions, a preferential leaching of U was observed with Zr-rich compositions. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). From these findings, dissolution with HNO$$_{3}$$ is expected to be only applicable in U-rich compositions ($$x$$ $$<$$ 0.3) if the dissolution in 6 M HNO$$_{3}$$ at 353 K is assumed. Application of complexing acids such as mixture of HNO$$_{3}$$ and HF should be considered to increase the dissolution rate of the phases with Zr-rich compositions.

論文

Suggestion of typical phases of in-vessel fuel-debris by thermodynamic calculation for decommissioning technology of Fukushima-Daiichi Nuclear Power Station

池内 宏知; 近藤 賀計*; 野口 芳宏*; 矢野 公彦; 鍛治 直也; 鷲谷 忠博

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1349 - 1356, 2013/09

For the decommissioning of Fukushima-Daiichi Nuclear Power Station (1F), characterization of fuel-debris in cores of Unit 1-3 is necessary. In this study, typical phases of fuel-debris generated in reactor pressure vessel were suggested by means of thermodynamic calculation using compositions of core materials and core temperatures. At low ogygen potential where metallic zirconium remains, (U,Zr)O$$_{2}$$, UO$$_{2}$$, and ZrO$$_{2}$$ were formed as oxides, and oxygen-dispersed Zr, Fe$$_{2}$$(Zr,U), and Fe$$_{3}$$UZr$$_{2}$$ were formed as metals. With an increase in zirconium oxidation, the mass of those metals, especially Fe$$_{3}$$UZr$$_{2}$$, were decreased, but the other phases of metals hardly changed qualitatively. Consequently, (U,Zr)O$$_{2}$$ is suggested as a typical phase of oxide, and Fe$$_{2}$$(Zr,U) is suggested as that of metal. This result can contribute to the characterization of debris in 1F, which will be also revised by considering the effect of iron content in RPV.

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

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

池内 宏知; 佐野 雄一; 柴田 淳広; 小泉 務; 鷲谷 忠博

Journal of Nuclear Science and Technology, 50(2), p.169 - 180, 2013/02

 被引用回数:5 パーセンタイル:53.31(Nuclear Science & Technology)

高速炉燃料再処理において高重金属濃度の溶解液を得るため、燃料の短尺せん断片を用いた効率的な溶解プロセスを確立した。照射済MOX燃料の溶解速度に対するPu富化度,重金属濃度、及びせん断長さによる影響を調査した。その結果、照射済燃料の溶解速度はPu富化度の上昇とともに指数関数的に減少するが、未照射燃料と比べて100から1000倍程度増加することがわかった。浸透理論による速度論的解析の結果、重金属濃度の増加により固液比が減少し、燃料への硝酸の浸透性が低下することで溶解速度が低下することが示唆された。せん断長さの短尺化(長さ10mm)により、燃料の比表面積が増加することで硝酸の浸透性が改善され、高重金属濃度の条件においても従来再処理における濃度条件と同程度に溶解速度が維持されることがわかった。

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

FaCT Phase-I evaluation on the advanced aqueous reprocessing process, 3; Highly effective dissolution technology for FBR MOX fuels

池内 宏知; 桂井 清道*; 佐野 雄一; 鷲谷 忠博

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Japan Atomic Energy Agency (JAEA) has been developing an efficient dissolution technology for irradiated MOX fuel in the framework of Fast Reactor Cycle Technology Development (FaCT) Project. In the fuel dissolution process for advanced aqueous reprocessing system named NEXT (New Extraction System for TRU Recovery), highly concentrated dissolution is required to adapt to the crystallization process. Optimum dissolution condition including short stroke shearing or pulverization of the irradiated fuel has been discussed, being based on the calculation results of continuous dissolver simulation code which reflects the results of dissolution tests using irradiated MOX fuel. We have been also developing rotary drum type continuous dissolver to adapt to the dissolution process for high heavy metal (U and Pu) concentration. This paper describes the summary and evaluation of R&D results on highly effective dissolution technology in the framework of FaCT Phase-I from 2006 to 2010.

論文

FaCT Phase I evaluation on the advanced aqueous reprocessing process, 5; Research and development of uranium crystallization system

柴田 淳広; 矢野 公彦; 三本松 勇二; 中原 将海; 竹内 正行; 鷲谷 忠博; 長田 正信*; 近沢 孝弘*

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

FaCTプロジェクトの一環として、ウラン晶析技術開発を実施している。開発目標は、70%以上のU回収率,100以上のDF,晶析装置の性能確認などである。実溶解液を用いたビーカ規模試験により基礎データを取得した。U晶析率は供給液組成や冷却温度により制御可能である。大半のFPのDFは洗浄操作により改善する。しかしながら、Pu-Cs複塩の生成によりCsの低DFが生じている。円環型晶析装置及び結晶分離機の性能を確認するため、各種試験を実施し、良好な機械的性能を確認した。しかしながら、結晶分離機によるU結晶の洗浄は、固体不純物に対して効果が認められなかった。U結晶の純度を改善するため、結晶精製技術の導入を検討し、KCP(Kureha Crystal Purifier)を選定した。KCPにおける固体不純物の挙動把握のため、ベンチスケールのKCP装置を用いてU結晶精製試験を実施した。KCPは液体不純物のみならず、固体不純物についてもよい除染性能を示した。

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