A new technology for the nuclear industry for the complete and continuous pyrochemical reprocessing of spent nuclear fuel: Catalyst enhanced molten salt oxidation / Griffiths Trevor R.,Volkovich Vladimir A. // NUCLEAR TECHNOLOGY. - 2008. - V. 163, l. 3. - P. 382-400.

ISSN/EISSN:

0029-5450 / нет данных

Type:

Article; Proceedings Paper

Abstract:

With the current and forthcoming need to develop new nuclear power plants, decommission existing nuclear plants, and satisfy future demands to minimize nuclear waste, it is important to examine and test potential new technologies instead of limit the methods to the original techniques. The modern and future safeguards that are or will be imposed on the industry will be more restrictive than in the early years, and it behooves the nuclear industry to consider and employ recent beneficial developments. An approach to a complete pyrochemical reprocessing cycle-which can be closed and nonaqueous-that employs catalyst enhanced molten salt oxidation (CEMSO) is outlined. We believe that this proposed new technology is faster and has the potential to be more complete than the two main existing technologies, will produce fission products in a compact and suitable form for vitrification, and will have additional cleanup and other advantages. An outline of our process was presented at the Seventh International Symposium on Molten Salts Chemistry and Technology in Toulouse, France, and an invitation was extended for a longer account. We thus present a sufficient description of our process (developed at a university not at a nuclear institute, and based on the data in our publications) for it to be taken to pilot-plant scale, without initially employing highly radioactive isotopes, together with all our background data, or references thereto, obtained over a 10-yr period. To assist readers, the titles of our 50+ publications and those of others are given in the reference section. CEMSO is a technology that we are convinced has much future promise. This account has been written with reprocessing and nuclear engineers and technologists in mind and is aimed at helping them understand the potential and interesting subtleties of molten salt chemistry, a topic the), will not have previously encountered. The reasons why the original molten salt oxidation (MSO) experiments on nuclear fuel and waste in the last century were limited and discontinued are here shown to arise from a misunderstanding of the chemistry involved. Thus, the full potential and advantages for a rapid and efficient separation and recycling technology, which we have established, were missed. Uranium (and plutonium) in spent fuel can be converted to insoluble uranates by air sparging in molten carbonates (solubility of uranates similar to 200 ppm). (Data on plutonates are not determined but are expected to be similar.) The fission product elements remain in solution and are concentrated and later precipitated (>97\% efficiency) as phosphates, the carbonate melt is recycled, and minor gaseous products can be trapped. Areas needing further research, impossible for us, are mainly the chemistry of transuranic actinides in molten carbonates and of their phosphate compounds and the design and manufacture of pilot-plant test rigs. We have investigated the 24 major fission product elements Cs, Mg, Sr, Ba, lanthanides (La to Dy), Zr, Cr, Mo, Mn, Re (to simulate Tc), Fe, Ru, Ni, Cd, Bi, and Te in molten chlorides, considered comparable and more convenient than carbonate. Cesium was removed completely as a complex phosphate precipitate. Melts containing several fission product elements were explored for possible coprecipitation reactions. The previously developed industrial technology for MSO can be employed with CEMSO to create in minimum time a fast, efficient technique for reprocessing nuclear fuel and concentrating fission products and radioactive waste, applicable to current decommissioning and future nuclear reactor technology.

Author keywords:

phosphate precipitation; molten carbonate; fission products ALKALI-METAL URANATES; CHLORIDE MELTS; FISSION-PRODUCTS; SODIUM-CARBONATE; URANIUM-DIOXIDE; ABSORPTION-SPECTROSCOPY; GRAPHITE OXIDATION; OXYGEN SOLUBILITY; UO2; MIXTURES

DOI:

нет данных

Web of Science ID:

ISI:000258910700007

Соавторы в МНС:

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Другие поля

Поле	Значение
Month	SEP
Note	Symposium on Molten Salt Chemistry and Technology, Toulouse, FRANCE, AUG, 2005
Publisher	AMER NUCLEAR SOC
Address	555 N KENSINGTON AVE, LA GRANGE PK, IL 60526 USA
Language	English
Keywords-Plus	ALKALI-METAL URANATES; CHLORIDE MELTS; FISSION-PRODUCTS; SODIUM-CARBONATE; URANIUM-DIOXIDE; ABSORPTION-SPECTROSCOPY; GRAPHITE OXIDATION; OXYGEN SOLUBILITY; UO2; MIXTURES
Research-Areas	Nuclear Science \& Technology
Web-of-Science-Categories	Nuclear Science \& Technology
ResearcherID-Numbers	Volkovich, Vladimir/K-8339-2012
ORCID-Numbers	Volkovich, Vladimir/0000-0003-4438-1194
Number-of-Cited-References	77
Usage-Count-Since-2013	15
Journal-ISO	Nucl. Technol.
Doc-Delivery-Number	344EZ