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Title: Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

Abstract

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.

Inventors:
; ; ;
Publication Date:
Research Org.:
Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1399866
Patent Number(s):
9,793,023
Application Number:
14/042,115
Assignee:
Los Alamos National Security, LLC LANL
DOE Contract Number:
AC52-06NA25396
Resource Type:
Patent
Resource Relation:
Patent File Date: 2013 Sep 30
Country of Publication:
United States
Language:
English
Subject:
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CHEMISTRY; 36 MATERIALS SCIENCE

Citation Formats

Reilly, Sean Douglas, May, Iain, Copping, Roy, and Dale, Gregory Edward. Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target. United States: N. p., 2017. Web.
Reilly, Sean Douglas, May, Iain, Copping, Roy, & Dale, Gregory Edward. Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target. United States.
Reilly, Sean Douglas, May, Iain, Copping, Roy, and Dale, Gregory Edward. 2017. "Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target". United States. doi:. https://www.osti.gov/servlets/purl/1399866.
@article{osti_1399866,
title = {Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target},
author = {Reilly, Sean Douglas and May, Iain and Copping, Roy and Dale, Gregory Edward},
abstractNote = {A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2017,
month =
}

Patent:

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  • Molybdenum is separated from molybdenum-containing activated charcoal or char also containing small amounts of uranium obtained as a by-product in uranium leaching processes by stripping with an alkaline solution to provide a molybdenum containing solution containing substantially less than 500 ppm u/sup 3/o/sup 8/.
  • A process for the separation and collection of molybdenum-99 from an irradiated uranium-containing target material utilizes thermal chromatographic separation. The irradiated target material containing the molybdenum-99 is heated in an oxidizing atmosphere to form an oxidized target material and gaseous molybdenum-99 trioxide. The gaseous molybdenum-99 trioxide is carried by the oxidizing atmosphere along with other vaporized materials to a cooling zone for progressive condensation and collection of the molybdenum-99 trioxide and the other materials in the form of separate deposits.
  • ln processes for extracting U from its ores using liquid amine extractants, Mo has caused some operating difficulty in that it often precipitates as a complex with the amine in the process and interferes with the efficient separation of the U. A method for circumventing this difficulty is given in which the amine extractant containing the U and Mo is scrubbed with a dilute aqueous solution of H/sup 2/O/sup 2/ to maintain Mo in the molybdate state. The treated extractant is then stripped by an aqueous NaCl solution. (D.L.C.)
  • A process for recovering uranium values from a sulphate solution containing dissolved uranium and molybdenum and with a pH not exceeding about 5.5, includes reacting the solution with ammonia at a pH in the range of from about 8 to about 10, without the solution existing for any significant time at a pH of around 7, with resultant precipitation of uranium values relatively uncontaminated by molybdenum. The uranium containing precipitate is separated from the remaining solution while maintaining the pH of the remaining solution within the same range.