Recovering Uranium from Graphite Fuel Elements

Mellitic Acid from the Oxidation of Graphite with 90% Nitric Acid.

“…: Caled., 70.0; Found, 71.3. Pyromellitic (1,2,4, 5-benzene tetracarboxylic) acid: Eastman Organic Chemicals practical grade (greater than 90%). Anal.…”

“…SeVERAL METHODS utilizing nitric acid for the processing of graphite-matrix nuclear reactor fuels have been studied (1,2,5). Generally, 4 to 15M HN03 is employed; however, in one of these procedures (2) the fuel is simultaneously disintegrated and leached with boiling 90% HN03.…”

“…SeVERAL METHODS utilizing nitric acid for the processing of graphite-matrix nuclear reactor fuels have been studied (1,2,5). Generally, 4 to 15M HN03 is employed; however, in one of these procedures (2) the fuel is simultaneously disintegrated and leached with boiling 90% HN03. Prior studies (3, 7, 9) showed that soluble species such as oxalic and mellitic (benzene hexacarboxylic) acids were produced in the oxidation of graphites with various aqueous reagents.…”

Dissolution of U-Mo-Alloy Scrap from Fuel Fabrication

“…In this process, the fuel is simultaneously pulverized, and uranium is leached in 90% HNO 3 . The uranium losses are greater then 1% when uranium content in graphite based fuel is 1.5% (w/w) and 0.15% when the uranium content is 5% (Bradley & Ferris 1961).…”

Aqueous Processing of Thorium Fuels

“…Graphite-base Fuels, Three potential methods for processing graphite-base fuels are being developed: grind-leach (~70% HNO3) (39), disintegration-leach (90%HNO3) (40), and combustion-dissolution. While the grind-leach process is applicable to virtually all types of graphite-base fuels proposed, uranium and thorium losses to the undissolved matrix are generally higher than desired.…”