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Journal articles on the topic 'Pyrochemical treatment'

Author: Grafiati

Published: 31 May 2025

Last updated: 31 July 2025

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1

Hayashi, Hirokazu, Takumi Sato, Hiroki Shibata, Masaki Kurata, Takashi Iwai, and Yasuo Arai. "Pyrochemical treatment of spent nitride fuels for MA transmutation." Science China Chemistry 57, no. 11 (2014): 1427–31. http://dx.doi.org/10.1007/s11426-014-5205-9.

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2

Harrison, M. T., and D. McKendrick. "Treatment of waste salt arising from the pyrochemical treatment of used nuclear fuel using precipitation methods." MRS Advances 7, no. 5-6 (2022): 117–21. http://dx.doi.org/10.1557/s43580-022-00244-z.

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3

Krachler, Michael, Rafael Alvarez-Sarandes, Pavel Souček, and Paul Carbol. "High resolution ICP-OES analysis of neptunium-237 in samples from pyrochemical treatment of spent nuclear fuel." Microchemical Journal 117 (November 2014): 225–32. http://dx.doi.org/10.1016/j.microc.2014.07.005.

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4

Gilbert, M. R. "Pressureless Sintering of Sodalite Waste-forms for the Immobilization of Pyroprocessing Wastes." MRS Proceedings 1744 (2015): 61–66. http://dx.doi.org/10.1557/opl.2015.394.

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ABSTRACTSodalite (Na8[AlSiO4]6Cl2), a naturally occurring Cl-containing mineral, has long been regarded as a potential immobilization matrix for the chloride salt wastes arising from pyrochemical reprocessing operations, as it allows for the conditioning of the waste salt as a whole without the need for any pre-treatment. Here the consolidation and densification of Sm-doped sodalite (as an analogue for AnCl3) has been investigated with the aim of producing fully dense (i.e. > 95 % t.d.) ceramic monoliths via conventional cold-press-and-sinter techniques at temperatures of < 1000 °C. Micr

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5

N., K. MITRA, DAS D., SEN ALOKE, and BASUMAJUMDAR A. "Influence of Exchangeable Cations on DTA Characteristics of Kaolinite and Montmorillonite Clays." Journal of Indian Chemical Society Vol. 62, Sep 1985 (1985): 662–65. https://doi.org/10.5281/zenodo.6322175.

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 Department of Chemical Technology, University of Calcutta, Calcutta-700 009<strong> </strong> <em>Manuscript received 28 August 1984, revised 7 May 1985, accepted 11 September 1985</em> The influence of exchangeable cations in the structure of kaolinite and montmorillonite minerals on the phase transformation on progressive heat treatment, has been studied through DTA curves. Hydration energy of the cations influences the expulsion of water from the clay lattice. The high temperature phase conversions were also influenced catalytically by exchangeable cations. 

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6

KOTANI, Masaki, and Shinji OGIHARA. "Effects of pyrochemical treatment for producing interfacial carbon layer on tensile properties of PIP-processed SiC/SiC composites." Proceedings of the Materials and processing conference 2019.27 (2019): 422. http://dx.doi.org/10.1299/jsmemp.2019.27.422.

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7

Rodrigues, Davide, Gabriela Durán-Klie, and Sylvie Delpech. "Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step." Nukleonika 60, no. 4 (2015): 907–14. http://dx.doi.org/10.1515/nuka-2015-0153.

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Abstract The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves

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8

Kotani, Masaki, and Shinji Ogihara. "Effects of pyrochemical treatment for producing interfacial carbon layer on tensile properties of polymer-derived-matrix-based SiC/SiC composites." Journal of Asian Ceramic Societies 9, no. 2 (2021): 519–30. http://dx.doi.org/10.1080/21870764.2021.1896095.

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9

Martinot, L., and M. Ligot. "The potential/pO2− diagrams of uranium and the feasibility of a pyrochemical head-end treatment of nuclear fuels in molten chlorides." Journal of Radioanalytical and Nuclear Chemistry Letters 136, no. 1 (1989): 53–60. http://dx.doi.org/10.1007/bf02165086.

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10

Gonzalez, M., L. Hansen, D. Rappleye, R. Cumberland, and M. F. Simpson. "Application of a One-Dimensional Transient Electrorefiner Model to Predict Partitioning of Plutonium from Curium in a Pyrochemical Spent Fuel Treatment Process." Nuclear Technology 192, no. 2 (2015): 165–71. http://dx.doi.org/10.13182/nt15-28.

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11

Volkovich, Vladimir A., Trevor R. Griffiths, and Robert C. Thied. "Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts." Journal of Nuclear Materials 323, no. 1 (2003): 49–56. http://dx.doi.org/10.1016/j.jnucmat.2003.08.024.

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12

Fujita, Reiko. "(Keynote) Pyrochemical Process in Molten Salts for Spent Nuclear Fuel Reprocessing and Radioactive Waste Treatments." ECS Transactions 86, no. 14 (2018): 311–20. http://dx.doi.org/10.1149/08614.0311ecst.

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13

Harrison, Mike T., Howard E. Simms, Angela Jackson, and Robert G. Lewin. "Salt waste treatment from a LiCl-KCl based pyrochemical spent fuel treatment process." Radiochimica Acta 96, no. 4-5 (2008). http://dx.doi.org/10.1524/ract.2008.1490.

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Spent nuclear fuel may be treated using molten salt electrochemical techniques to separate fission products and actinide metals. Salt waste arising from the electrorefining process contains alkali metals, alkaline-earth and rare earth fission products, along with residual actinides. The removal of fission product elements has been investigated using zeolite ion exchange and phosphate precipitation, which allow the salt electrolyte to be recycled back into the main electrorefining vessel. Recycling the salt minimizes the volume of high level waste (HLW) generated and yields the fission products

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14

Liu, Xueyang, Xilei Duan, Lin Li, Qiang Zhang, Zhenghua Qian, and Yanbo Qiao. "Synthesis and characterization of cerium monazite phosphate glass‐ceramic for immobilization of nuclear waste salt." International Journal of Applied Ceramic Technology, October 2024. http://dx.doi.org/10.1111/ijac.14940.

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AbstractA phosphate glass‐ceramic wasteform incorporating cerium monazite was synthesized through a two‐step solid phase sintering methodology, aimed at immobilizing waste salts derived from the pyrochemical reprocessing of spent nuclear fuel. The initial phase involved the synthesis of cerium monazite powder, employing a stoichiometric mixture of CeF3 and NH4H2PO4 in a molar ratio of mol(Ce:PO4) = 1:1.4, which was subjected to a thermal treatment at 900°C. Subsequently, the glass‐ceramic matrix was fabricated by sintering a composite mixture of the synthesized cerium monazite and iron phospha

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15

"(Keynote) Pyrochemical Process in Molten Salts for Spent Nuclear Fuel Reprocessing and Radioactive Waste Treatments." ECS Meeting Abstracts, 2018. http://dx.doi.org/10.1149/ma2018-02/53/1886.

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