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G. R. JASNY, J. R. BARKMAN, J. P. SPRAGUE, and R. P. SMITH Union Carbide Nuclear Co., Oak Ridge, Tenn.
Recovering Uranium from Unirradiated Fuel Element Scrap
Ma,
development work in nuclear fuel recovery has been devoted to recovering irradiated fuel elements, the reasoning being that what is good for irradiated fuel elements must be better for unirradiated fuel. This is basically sound, but to understand the over-all recovery problem, the difference between irradiated fuel elements and unirradiated fuel fabrication scrap must be appreciated. Unirradiated enriched uranium fuel fabrication scrap is any material generated in fabricating enriched uranium fuel elements, which contains enough enriched uranium for economic recovery. Because of the high value of enriched uranium (77),it ranges from bona fide scrap such as metal turnings and casting dross to dirty rags and mop water. Dealing with this type of scrap involves four recovery problems: nuclear safety, health protection, accountability, and nonuniformity of shape, concentration, and contaminants. Uranium concentrations of scrap are usually poorly known; therefore batchwise control is unwise, Safe geometry processing equipment must be relied on, which means tubular processing vessels, 6 inches in diameter or less, slabs 2 inches thick, or volume-limited 4-liter containers. I n special cases, such as burning or crushing, dryness is relied on for nuclear safety. Health protection is simplified because no appreciable external radiation is involved. Inhalation and ingestion of uranium and its compounds must be prevented; therefore, most recovery work must be done in dry boxes, hoods, and closed vessels.
This report on a neglected (and expensive) part of the uranium fuel cycle shows solution techniques for U-Zr, U-stainlesssteel, and &AI alloy scrap recovery of enriched uranium by dibutyl Carbitol extraction, with 99.97” over-all efficiency
A large fraction of unirradiated scrap most common forms of uranium-stainless is shipped to a central recovery plant. alloy are metal strips l/l6 to l/g inch Unfortunately, “by difference” accountthick, 2 to 4 inches wide, and varying in ing on the part of the shipper is usually length up to 15 inches. Uranium coninevitable, and processors sometimes centrations vary from 3 to 15%. have two conflicting responsibilities: The bulk of the alloy is electrodissolved They must select a sampling point in the in a nitric acid electrolyte (Figure 2). recovery process which permits early Alloy “end pieces,” held in the anode blending with other streams, so that clamps during electrodissolution, are then inefficient paralleling of facilities can be dissolved in boiling hydrofluoric acid. minimized, and sampling and subsequent The electrodissolution produces a soluanalytical work must establish agreetion of nitrates of iron, nickel, chromium, ment with the shipper or, in the event of ’ and uranium, as well as hydrogen and disagreement, convince the shipper #that mixed oxides of nitrogen which are the processor’s reckoning is correct. vented. Niobium, if present, is not affected and must be filtered out. Nonuniformity of unirradiated scrap Hydrofluoric acid (15 M ) although relaaffects processing efficiency. The wide tively slow in its dissolving action, is used range of size distribution is combined to dissolve the end pieces because of the with the limitations of safe geometry subsequent ease of complexing the fluoequipment to require the frequent use of ride ions with aluminum nitrate to small batch equipment in the dissolving produce a solution, like the electrooperations. Contaminants of every kind dissolver solution, compatible with the require careful spot checking prior to stainless steel equipment used in extracprocessing; wide fluctuations in uranium concentration complicate preparation of tion. uniform extraction feed. The electrodissolution is carried out at Chemically, the recovery methods 80” to 90” C. and requires 5 to 6 volts. used (Figure 1) parallel those in aqueous The alloy anodes are supported in the reprocessing of corresponding irradiated electrolyte, 4M nitric acid, by stainless fuel elements (76). steel clamps. The stainless steel dissolving vessels are the cathodes. Rate of dissolution i s controlled by Dissolution regulating the current and specific Uranium-Stainless Steel Alloys. The gravity of the solution. As the alloy is J
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Aqueous dissolution
1-1
u-SS Electrodissolution in ”03
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U-Zr Dissolution in HF-“01
U-A1 Dissolution in NaOH-NaN03-Ba(NOa)2
Addition of Al(N0s)s and acidity adjustment
Feed preparation 1
1
I
Primary purification and recovery Concentration
Evaporation
Final purification
c
Pure uranyl nitrate Figure 1. Summary flowsheet of the over-all enrkhed uranium scrap dissolving and purification process VOL. 50, NO. 12
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DECEMBER 1958
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dissolved, the anode clamps are lowered, exposing additional alloy to the solution. The end pieces may be reduced to halfinch squares through successive readjustments. When the concentration of the solution reaches a specific gravity of 1.32 to 1.34, the solution is changed. The uranium-bearing solution recovered from the electrodissolvers contains 2 to 10 grams of uranium per liter, depending on the uranium content of the alloy. The electrodissolving installation utilizes slab and cylindrical geometry. The rectangular tanks are 2 inches wide, 24 inches high, and 24 inches long; the cylindrical tanks are 6 inches in diameter and 24 inches high. Direct current is provided from two copper oxide rectifiers, each rated a t 1400 amperes and 6 volts, which feed all the dissolvers in parallel. At full capacity this installation, consisting of six rectangular and four cylindrical units, requires approximately 1200 amperes. Its average throughput is 2 kg. of alloy per hour. Uranium-Zirconium Alloys. Forms of uranium-zirconium alloy to be dissolved are milling machine scrap, lathe turnings, fines, and massive rods. Uranium concentrations range from 4 to 20%, and usually 1 to 2% of tin is present. This scrap is packed under hydrocarbon oil during shipment and storage. The alloy is degreased with trichloroethylene prior to dissolution. The dissolution method is very similar to that reported by Argonne National Laboratory (79). The alloy is dissolved in a slight excess of hydrofluoric acid and a small amount of nitric acid added a t the end of the reaction to dissolve the tin and oxidize the uranium to the uranyl ion. This method is relatively fast and results in a concentrated, stable solution. PROCESS.The alloy is dissolved according to the following reactions: U
+ 4HF + UFa + 2H2
Zr
+ 4HF
+
ZrFa f 2Hz
At least 4 moles of fluoride are required per mole of zirconium for safety as well
U-SS ALLOY
AN TO SAMPLING AND EXTRACTION
DISSOLVING CONDITIONS CURRENT DENSITY-0.2 amR/cm? VOLTAGE 4-6 volts DISSOLUTION RATE-0.25 gmdamp.hr. OPERATING TEMP.-80°-900c. ELECTROLYTE 4M HNO3 ANODE HOLDER SS347 clamp ' CATHODE SS347tank
-
CENTRIFUGE Figure 2. Flowsheet shows the electrodissolution process used on uraniumstainless steel alloy scrap
~
as stoichiometry reasons: Explosive reactions will not occur if 4 moles of fluoride are present for each mole of zirconium (72). If the fluoride-zirconium ratio (70) exceeds 5, insoluble uranium tetrafluoride precipitates. If tin is present in the alloy, it is also insoluble in the hydrofluoric acid solution, but it readily dissolves upon heating with twice the theoretical amount of nitric acid. The uranium is oxidized to uranyl nitrate by the nitric acid and precipitated uranium tetrafluoride will dissolve. Approximately 500 grams of alloy are placed in a graphite vessel with 2 liters of water, and 750 ml. of 52% hydrofluoric acid are added a t a rate to maintain a rapid reaction. Massive alloy permits fast addition of the acid, while fines require a trickle because of the violence of the reaction. Approximately 4.5 to 5.0 moles of fluoride are added per mole of zirconium. After the reaction has stopped, 100 ml. of 12M nitric acid are added. The solution is boiled for
Separation Factors between Uranium and Other Elements with Dibutyl Carbitol
(741 Organic 0.8ill
Ag
AI Ba B Be Ca
D
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=
H+
8 X 103
7 x 106 15 X 106 3 x 104 2
x
106
5 x 108
Corgmio/Caqueow.
INDUSTRIAL AND ENGINEERING CHEMISTRY
-
Aqueous Principally Mg(N03)2 9.9iM nitrate 0.13M H +
Separation Factors = D a u r a n i u m / D i m p u n t y Cr 5 x 105 Mg cu 2 x 106 Mo Fe 4 x 104 Na K 2 x 106 Ni Li 1 x 106 Sn Zr (F complex) Mn 2 x 106
several minutes until the tin is dissolved, the uranium is oxidized to the uranyl ion, and the excess hydrofluoric acid driven off, resulting in a clear solution. After 2 to 4 hours, depending on the form of alloy, the solution is filtered and placed in a safe geometry container. The uranium remaining in the residue has ranged from 0.01 to 0.5% of the total and averages about 0.25%. Accountability is established on the basis of analysis of dissolver solution and residues. Average uranium recovery has been 99.95%. Uranium-Aluminum Alloy. Although uranium-aluminum alloy is generally dissolved with nitric acid, using mercuric ions as a catalyst ( 6 ) , caustic dissolution has been considered (3, 73) as a tantalizing possibility. Caustic processing preferentially dissolves aluminum and silica. The uranium is oxidized to a thick concentrated sludge which is separated from the solution and redissolved in a small volume of acid; the filtrate is discarded. The major problems are separation of the sludge from the caustic filtrate, and elimination of soluble forms of uranium, particularly the uranyl carbonate complex. The basic reaction ( 3 ) for caustic dissolution is 2A1 f 2NaOH 4- 2Hz0 2NaA102 -b 3Hz (1) However, addition of sodium nitrate suppresses formation of hydrogen ( 3 ) , as shown in Equations 2 and 3. SA1 5NaOH 3NaN03 42H20 + 8NaA102 4- 3NH3 (2) 2A1 2NaOH 3NaNOa 4 2NaA10z 3NaNOz HzO (3) The uranium is precipitated as in-
5
x 106
9 x 102 9 x 104 14 X 106 2 x 105 2 x 103
+ +
+ + +
+
NUCLEAR TECHNOLOGY soluble oxides and uranates, which may be separated by centrifuging. However, the carbonate ions in the dissolver feed solution and those generated by the absorption of carbon dioxide from the atmosphere form a soluble uranyl carbonate ion (7), probably U O Z ( C O ~ ) ~ - ~ , which can result in relatively high uranium losses in the filtrate and increased recovery costs, as the filtrate would have to be processed through the extraction system. This is remedied by addition of barium nitrate, which precipitates carbonate ions as insoluble barium carbonate ( 3 ) . The scrap is received as billets, turnings, strips, and fines. The billets have to be reduced in size for processing in restricted geometry equipment, by recasting into slugs l’/z inches in diameter. However, a large fraction of this scrap cannot be recast because of the presence Figure 3. The caustic dissolution process is used on uranium-aluminum alloy scrap of excessive amounts of silica and carbon. These contaminants create problems C4Hg-O-CzH4-O-CzH4-O-CqHg. with alloy. A conical perforated screen further on in the process. Carbitol has a boiling point of 254.6’ C., a t the bottom of the column retains the The dissolution process is summarized a flash point of 118’ C., a specific metal in the column, but permits the in Figure 3. The scrap as received or gravity of 0.885, and a viscosity of slurry to pass with minimum plugging. recast is charged periodically into the 2.4 centipoises. I t gives excellent puriThe slurry flows by gravity through a top of a trickle-type column dissolver fication from most elements and very jacketed cooler to the centrifuge, elimi(2, 6). The caustic feed is prepared by good uranium recovery. nating the need for a slurry pump. dissolving barium oxide in nitric acid Zirconium is one of the more difficult Sharples Super Centrifuges, which rotate and adding sodium nitrate, sodium elements to separate, although the a t 15,200 r.p.m. and develop approxihydroxide, and water to produce a separation factor is still very high. mately 12,000 G’s, are used for all liquidsolution 3.6M in sodium hydroxide, Therefore, the purification and recovery solid separations. 2.5M in sodium nitrate, and 0.05M process used on a feed consisting of in barium nitrate. This solution, prezirconium-uranium dissolver solution Uranium Purification heated to 200’ F., is metered at the rate of blended with miscellaneous solutions has 15 gallons per hour into the top of the been singled out. The flowsheet (Figure The uranium may be recovered from column and trickles by gravity over the solutions by dibutyl Carbitol (Butex) 4) includes feed adjustment, first-cycle alloy. extraction. Dibutyl Carbitol (referred to extraction, evaporation, and secondThe slurry containing uranium oxides here as Carbitol) has the formula: cycle extraction. and insoluble contaminants is washed out of the bottom of the column by a FEED FIRST CYCLE EVAPORATION SECOND CYCLE slip stream of water and flows by gravity ADJUSTMENT EXTRACTION EXTRACTION into a high speed batch centrifuge. The filtrate is pumped to a polishing j < centrifuge of the same type, sampled for accountability, and discarded. The centrifuge cake is dissolved in 13M nitric acid in a column-type batch dissolver and the acid-insolubles, largely granular carbon, are removed in another centrifuge. The resulting concentrated uranyl nitrate solution is as much as 100 times more concentrated than the original caustic dissolver product. The residual solids are muffled at 600’ C. under oxygen and releached. Accountability is established by analysis of the concentrated uranyl nitrate solution, caustic filtrate, and muffled solids. Recovery has been between 99.8 and 99.9% of the uranium in the alloy, despite the discarding of over 90% of the caustic fil-AQUEOUS FLOW 2.5 trate. ----CAREITOL FLOW -6-VOLUMES The caustic dissolver consists of an RAFFINATE n t-nr PPMLI 8-foot length of 6-inch stainless steel pipe. The blank flange a t the top contains the liquid feed nozzle and is Figure 4. The dibutyl Carbitol process is used to remove and purify uranium hinged to permit charging the dissolver from the various uranium alloy dissolver solutions &’
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Feed adjustment consists of bringing the zirconium-uranium fluoride solutions to a specific gravity of 1.36 with aluminum nitrate and reducing the nitric acid concentration to 0.5M with sodium hydroxide semicontinuously under agitation in a long tube. The tube is kept full, unadjusted feed and chemicals are added in small batches, and adjusted feed is withdrawn continuously. Solutions from other alloy dissolutions as well as solutions such as pickling liquors and ash leachates are usually added at this point. I t has been determined at Argonne National Laboratory (70) and substantiated a t the Y-12 Plant that zirconium feed solutions have a rather narrow band of stability, as excess fluoride or aluminum will cause precipitation. Attempts to process feeds with too little or no fluorides usually result in disaster. The feed is stable until it reaches the extraction column, where extreme emulsification and precipitation occur and the system becomes inoperable. A high nitrate salt concentration in the aqueous feed is essential for good uranium recovery. and an acid concentration less than I N is preferred (5. 8, 9, 78). Uranium distribution coefficients of 20 to 50 are usually encountered in plant operations. although thev may be much higher. Zirconium is not a good salting agent by itself; so aluminum nitrate is added. Aluminum nitrate also complexes the fluoride so that the solution can be handled in stainless steel equipment. The second-cycle raffinate and firstcycle scrub raffinate are blended with the feed and pumped through the first-cycle extraction columns countercurrent to the Carbitol. Additional zirconium is scrubbed from the Carbitol by a fluoride solution arising from the dissolution of other scrap. A fluoride solution is more effective in scrubbing zirconium than aluminum nitrate alone (75),presumably because the zirconium-fluoride complex ions are less extractable into the solvent.
Table II. Average Second-Cycle Extraction Product Analysis (Uranium concentration, 130 g./l.) Because of the diversity of aqueous feeds, conditions must be reported as averages
Impurity Ag A1 B
Ba Be Ca Cr cu Fe
P.P.M.O
