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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:
; ; ;
Issue Date:
Research Org.:
Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1399866
Patent Number(s):
9793023
Application Number:
14/042,115
Assignee:
Los Alamos National Security, LLC
Patent Classifications (CPCs):
G - PHYSICS G21 - NUCLEAR PHYSICS G21C - NUCLEAR REACTORS
G - PHYSICS G21 - NUCLEAR PHYSICS G21G - CONVERSION OF CHEMICAL ELEMENTS
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. Tue . "Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target". United States. 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 = {Tue Oct 17 00:00:00 EDT 2017},
month = {Tue Oct 17 00:00:00 EDT 2017}
}

Works referenced in this record:

Production of High Purity Radioactive Isotopes
patent, May 1968


Production of High Purity Fission Product Molybdenum-99
patent, March 1974


Process for concentrating technetium-99m
patent, November 1979


Method of separating fission molybdenum
patent, April 1996


Medical isotope production reactor
patent, January 1997


Method of recovering uranium and transuranic elements from spent nuclear fuel
patent, March 2000


Process for reprocessing spent nuclear fuels by utilizing phenomenon of cocrystallization
patent, March 2006


Processes for recovering metals from aqueous solutions
patent, December 2012


Method of producing molybdenum-99
patent, May 2013


Process for reprocessing spent nuclear fuels by utilizing phenomenon of cocrystallization
patent-application, July 2003


Method and Apparatus for the Extraction and Processing of Molybdenum-99
patent-application, August 2011


Column Geometry to Maximize Elution Efficiencies for Molybdenum-99
patent-application, October 2011


Methods Of Separating Medical Isotopes From Uranium Solutions
patent-application, November 2012


Fuel Preparation For Use In The Production Of Medical Isotopes
patent-application, April 2014


Recovering And Recycling Uranium Used For Production Of Molybdenum-99
patent-application, March 2015


Study on the separation of molybdenum-99 and recycling of uranium to water boiler reactor
journal, January 1989

  • Cheng, W. L.; Lee, C. S.; Chen, C. C.
  • International Journal of Radiation Applications and Instrumentation. Part A. Applied Radiation and Isotopes, Vol. 40, Issue 4, p. 315-324
  • https://doi.org/10.1016/0883-2889(89)90224-4

Improved processes of molybdenum-99 production
journal, June 1999


Flowsheet Study of U-Pu Co-Crystallization Reprocessing System
journal, January 2012


Challenges of Extracting and Purifying Fission-Produced Molybdenum-99
journal, September 2000

  • McDonald, Marion J.; Carson, Susan D.; Naranjo, Gerald E.
  • Industrial & Engineering Chemistry Research, Vol. 39, Issue 9, p. 3146-3150
  • https://doi.org/10.1021/ie990377h

Large scale production of fission 99Mo by using fuel elements of a research reactor as starting material
journal, August 1984


Crystallization of Uranium Complexes for Partitioning of Spent Nuclear Fuel
book, June 2006


Precipitation Behavior of Dicesium Tetravalent Plutonium Hexanitrate in Cooling Crystallization of Uranyl Nitrate Hexahydrate
journal, February 2011


Purification Rate of Uranyl Nitrate Hexahydrate Crystal for Transuranium Elements on Isothermal Sweating Phenomenon
journal, November 2010

  • Nakahara, Masaumi; Nomura, Kazunori; Koizumi, Tsutomu
  • Industrial & Engineering Chemistry Research, Vol. 49, Issue 22, p. 11661-11666
  • https://doi.org/10.1021/ie1012097

Enhancement of Decontamination Performance of Impurities for Uranyl Nitrate Hexahydrate Crystalline Particles by Crystal Purification Operation
journal, April 2011


Behavior of Actinide Elements and Fission Products in Recovery of Uranyl Nitrate Hexahydrate Crystal by Cooling Crystallization Method
journal, April 2011


Effect of crystal size on purity of uranyl nitrate hexahydrate crystalline particles grown in nitric acid medium
journal, November 2012


Scrambling to Close the Isotope Gap
journal, January 2011


The isolation of 99Mo from fission material for use in the 99Mo/99mTc generator for medical use
journal, January 2004


Continuous-Operation Test at Engineering Scale Uranium Crystallizer System
conference, January 2009