Microparticles with diverse sizes and morphologies from mechanical and laser cutting of fuel debris simulants and geopolymer as a covering material

Zhou, Q.*; 斉藤 拓巳*; 鈴木 誠矢; 矢野 公彦; 鈴木 俊一*

Journal of Nuclear Science and Technology, 58(4), p.461 - 472, 2021/04

https://doi.org/10.1080/00223131.2020.1869623

During the decommissioning of Fukushima Daiichi Nuclear Power Plant, the dismantling of massive fuel debris is important for the defueling process. The production and dispersion of radioactive microparticles highly depends on the cutting technique implemented. Previous studies have been conducted on the development of cutting techniques and the treatment of radioactive particles generation during the cutting process. Besides, adequate understanding of the microparticle products during the cuttings of fuel debris is of vital necessity. Nowadays, geopolymer application is proposed as a covering material during the fuel debris retrieval in order to keep the structural integrity of damaged components. In this study, the microparticle products during mechanical and laser cutting of fuel debris simulants (Hafnium oxide and Tungsten(IV) oxide pellets) and geopolymer material were investigated. The cuttings have been carried out for samples of the simulants, geopolymer, and simulants with geopolymer covering. The generated particles were collected and investigated by laser diffraction particle size analyzer and Scanning Electron Microscope with X-ray spectrometry. Particles with diverse sizes and morphologies were observed from the products of each sample. It also appears that particles with unique sizes and morphologies can generate from the laser cuttings due to the thermal effect of laser cutting.

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

AA2018-0409.pdf:2.12MB

https://doi.org/10.1080/00223131.2019.1604272

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.

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

https://doi.org/10.15669/pnst.5.217

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, (U,Zr)O, 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.

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

https://doi.org/10.1016/j.jnucmat.2017.01.032

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 and (Zr,U)SiO, and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO requires a relatively long time. Therefore, the formation of (Zr,U)SiO 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.

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 content range from 10% to 65%.

Dissolution behavior of (U,Zr)O-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

https://doi.org/10.1080/00223131.2014.924445

To explore the possibility of dissolving fuel debris as a potential pre-treatment for waste treatment, dissolution tests of UZrO and (U,Pu)ZrO were carried out in 6 M HNO 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 is expected to be only applicable in U-rich compositions ( 0.3) if the dissolution in 6 M HNO at 353 K is assumed. Application of complexing acids such as mixture of HNO and HF should be considered to increase the dissolution rate of the phases with Zr-rich compositions.

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.

Nitric acid concentration dependence of dicesium plutonium(IV) nitrate formation during solution growth of uranyl nitrate hexahydrate

中原 将海; 鍛治 直也; 矢野 公彦; 柴田 淳広; 竹内 正行; 岡野 正紀; 久野 剛彦

Journal of Chemical Engineering of Japan, 46(1), p.56 - 62, 2013/01

https://doi.org/10.1252/jcej.12we175

U晶析工程においてCsPu(NO)の生成挙動に及ぼすHNO濃度の依存性を調べた。硝酸ウラニル溶液に対するCsPu(NO)の溶解度は、HNO濃度が高くなるほど低下する傾向を示した。照射済高速炉燃料溶解液を用いた晶析実験では母液のHNO濃度が6.5mol/dmの条件において、CsPu(NO)が析出し、硝酸ウラニル六水和物結晶に対するCsの除染係数は低下した。一方、母液のHNO濃度が4.0mol/dmのときは、CsPu(NO)が生成せず、硝酸ウラニル六水和物結晶とCsが良好に分離できることを示した。

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。

先進湿式法再処理の晶析工程におけるCs挙動把握のための模擬溶解液を用いた基礎試験

柴田 淳広; 矢野 公彦; 紙谷 正仁; 中村 和仁; 鷲谷 忠博; 近沢 孝弘*; 菊池 俊明*

日本原子力学会和文論文誌, 8(3), p.245 - 253, 2009/09

https://doi.org/10.3327/taesj.J08.050

U晶析工程におけるCsの挙動を調べるため、模擬溶解液を用いたU晶析バッチ試験及びU(IV)溶液を用いたCs複塩生成基礎試験を実施した。使用済燃料の溶解液中のCs濃度では、先進湿式法再処理のU晶析工程の条件においてCsNOやCsUO(NO)は生成せず、他のFP元素との相互作用によるCs塩も生成する可能性は小さいことを確認した。また、U(IV)溶液を用いたCs複塩生成基礎試験の結果から、酸濃度が5mol/dm以上の場合にはCsとPu(IV)の複塩が生成する可能性が示唆された。