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Title: Actinide removal from spent salts

Abstract

A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Inventors:
 [1];  [2];  [2];  [2];  [3]
  1. Pleasanton, CA
  2. Livermore, CA
  3. Danville, CA
Issue Date:
Research Org.:
Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
OSTI Identifier:
874838
Patent Number(s):
6471922
Assignee:
The Regents of the University of California (Oakland, CA)
Patent Classifications (CPCs):
C - CHEMISTRY C22 - METALLURGY C22B - PRODUCTION AND REFINING OF METALS
Y - NEW / CROSS SECTIONAL TECHNOLOGIES Y02 - TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE Y02P - CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
DOE Contract Number:  
W-7405-ENG-48
Resource Type:
Patent
Country of Publication:
United States
Language:
English
Subject:
actinide; removal; spent; salts; method; removing; contaminants; uranium; thorium; salt; molten; oxidation; mso; reactor; described; removed; analyzed; determine; carbonate; concentration; dissolved; water; reagents; added; precipitate; oxide; andor; diuranate; precipitated; materials; filtered; dried; packaged; disposal; radioactive; waste; 90; filtration; solutions; >20; returned; re-use; containing; <20; require; clean-up; exchange; column; yields; contain; 01; ppm; molten salt; radioactive waste; salt solution; /423/588/

Citation Formats

Hsu, Peter C, von Holtz, Erica H, Hipple, David L, Summers, Leslie J, and Adamson, Martyn G. Actinide removal from spent salts. United States: N. p., 2002. Web.
Hsu, Peter C, von Holtz, Erica H, Hipple, David L, Summers, Leslie J, & Adamson, Martyn G. Actinide removal from spent salts. United States.
Hsu, Peter C, von Holtz, Erica H, Hipple, David L, Summers, Leslie J, and Adamson, Martyn G. Tue . "Actinide removal from spent salts". United States. https://www.osti.gov/servlets/purl/874838.
@article{osti_874838,
title = {Actinide removal from spent salts},
author = {Hsu, Peter C and von Holtz, Erica H and Hipple, David L and Summers, Leslie J and Adamson, Martyn G},
abstractNote = {A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2002},
month = {1}
}

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