Nomenclature
Ajf
flow rate of aqueous stream entering contactor a t stage j A , = flow rate of aqueous stream leaving s t a g e j Y j = concentration of distributing component in organic stream leaving stage j X j = concentration of distributing component in aqueous stream leaving stage j 0,f = flow rate of organic stream entering contactor at stage j X,, = concentration of distributing component in aqueous stream entering contactor at stage j Y,f = concentration of distributing component in organic stream entering contactor a t stage j 0, = flow rate of organic stream leaving stage j Dj = equilibrium distribution coefficient Ar = number of stages in contactor X,~Q = equilibrium aqueous concentration of distributing component in stage j Y,eq = equilibrium organic concentration of distributing component in stage j =
Haas, W. O., “Chemical Processing of Reactor Fuels,” J. F. Flagg, ed., p. 126, Academic Press, New York, 1961. Haas, W. O., Znd. Eng. Chem. 50, 125 (1958). Hanson, D. N., Duffin, J. H., Somerville, G. F., “Computation of Multistage Separation Processes,” Reinhold, New York, 1962. Henry, H . E., “Isolating Americium and Curium from A1(K03)3NaNO--HNOq Solutions bv Batch Extraction with Tributvl Phosphgte,” E. I . du Pont de Nemours & Co., Savannah Rivdr Laboratory, USAEC Rept. DP-972 (1965). Mills, A. L., “Review of Computer Programmes for Solvent Extraction Calculations,” Reactor Group, United Kingdom Atomic Energy Authority, TRG Rept. 902 ( D ) (1965). Olander, D. R., Znd. Eng. Chem. 53, 1 (1961). Roth, J. A . , Henry, H. E., J . Chem. Eng. Data IO, 298 (1965). Schlea, C. S., Caverly, M. R., Horni, E. C., Henry, H. E., Jenkins, W.J., “Miniature Pilot Plant for Processing Irradiated Nuclear Fuel,” E. I. du Pont de Nemours & Co., Savannah River Laboratory, USAEC Rept. DP-757 (1962). Siddall, T. H., “A Rationale for the Recovery of Irradiated Uranium and Thorium by Solvent Extraction,” Proceedings of 2nd International Conference on Peaceful Uses of Atomic Energy, Vol. 17, p. 339, Geneva, 1958. RECEIVED for review January 20, 1967 ACCEPTEDOctober 12, 1967
literature Cited
Burton, TV. R., Mills, A . L., ,Vucl. Eng. 8, 248 (1963). Codding, J. T$‘., Haas, LV. O., Heuniann, F. K., Znd. Eng. Chem. 50, 145 (1958). DiLiddo, B. A , , IValsh, T. J., Znd. Eng. Chem. 53, 801 (1961). Eubanks, I. D., Burney, G. A . , “Curium Process Development. I. General Process Description,” E. I . du Pont de Nemours & Co., Savannah Riler Laboratory, USAEC Rept. DP-1009 (1966). Groh, €1. J., Huntoon, R . T., Schlea, C. S., Smith, J. A., Springer, F. H., .\‘uclear Appl. 1, 327 (1965).
Information developed during work under Contract AT(07-2)-1 with the U. S. Atomic Energy Commission. Material supplementary to this article has been deposited as Document No. 9793 with the AD1 Auxiliary Publications Project, Photoduplication Service, Library of Congress, IVashington, D. C. A copy may be secured by citing the docummt number and by remitting $6.25 for photoprints or 62.50 for 35-mm. microfilm. Advance pa! ment is required. Make checks or money orders payable to Chief, Photoduplication Service, Library of Congress.
SOLVENT EXTRACTION OF THORIUM AND URANIUM FROM BERYLLIUM NITRATE FEEDS BY TRI-n-BUTYL PHOSPHATE R. C. CAIRNS, M. G. B A I L L I E , B. J. FOX, A N D R . K. R Y A N Australian Atomic E n e r a Commission Research Establishment, Sutherland, ,Yew South Wales, Australia
A solvent-extraction process is presented for the recovery of uranium and thorium from radioactive feed solutions highly salted with beryllium nitrate. By using a tri-n-butyl phosphate extractant and a split contactor system, in which the scrub raffinate and feed raffinate streams are kept separate, it was demonstrated that fission product decontamination factors greater than 1 03,and adequate uranium and thorium recoveries, can be obtained for activity levels up to 1 curie per liter.
ABORATORY
development has recently been completed a t
L Lucas Heights of flo\ysheets for the fuel cycle of a B e 0
high temperature reactor system (Cairns et al., 1966). As part of this work it \yas necessary to develop a liquid-liquid extraction process for uranium and thorium recovery from feeds highly salted with beryllium nitrate. T h e fuel element (Smith, 1966) consists of a matrix of beryllium oxide, in the form of a sphere of about I-inch diameter, containing particles of plutonium and thorium oxides in solid solution, 150 to 200 microns in diameter. T h e sphere is coated with a iayer of beryllium oxide about 0.05 inch thick. Because of the presence of beryllium oxide this fuel element poses some unique extraction problems. Aqueous head-end methods, and liquid-liquid separation techniques, were selected
for experimental study and this led to the development of a solvent-extraction process for recovering uranium and thorium from the highly salted feed solution arising from the head-end step. B e 0 High Temperature Gas-Cooled Reactor Fuel Cycle
For economic reasons it was necessary to consider the recycle of both moderator and fuel material. The fuel cycle studies were therefore based on the conceptual flowsheet given in Figure 1. In the head-end process the fuel elements are crushed and ground, and subjected to a nitric acid leaching step to dissolve the bulk of the fuel materials with as little of the B e 0 moderator as possible; the residual moderator is subsequently dissolved for further treatment. In the actinide sepaVOL. 7
NO. 2 A P R I L 1 9 6 8
177
REACTORS
I
I FIRST REACTOR CORES
1
I BERYLLIUM STREAM
I RECYCLED REACTOR CORES
1
rRAFFINATE K I S S I O N PRODUCTS Pu (NO,),
PURIFICATION OF 0a STREAM AND
R E C O N V E R W TO B e 0 MAKE-UP Be 0
1
I
t
1
RECONVERSION OF ACTINIDES AND
REFABRICATION
PRODUCTON OF
FUEL PARTICLES
Figure 1.
Process steps in fuel cycle
ration step uranium and thorium are separated from the fission products and beryllium nitrate, by the use of a n organic solvent containing tri-n-butyl phosphate (TBP) as the extractant. While the fuel cycle studies proceeded, continual modifications to the fuel composition were being made. The most recent estimates of fuel inlet and outlet compositions for a 200-MWe.) high temperature gas-cooled reactor with fuel recycle, after 120 days' cooling, are given in Table I. Although these compositions are not final and neglect the buildup of higher isotopes of uranium in the fuel inlet composition, they served as a guide to experimental work. Development of the head-end process a t low activity levels indicated that, for satisfactory recovery of fuel materials, 20% of the beryllia present in the fuel would appear in the nitric acid leach solution. T h e conceptual solvent-extraction flowsheet for recovery and purification of actinides was based on this assumption, and it was also assumed that 99% recovery of thorium and 99.8% recovery of uranium-233 would be required. With the proposed fuel recycle scheme, decontamination factors from fission products of only lo3 were required, equivalent to that obtained from the first cycle of standard processes. In this work also, residual plutonium was assumed to be rejected to waste, since its isotopic composition after recycle would make it unsuitable for re-use in an H.T.G.C.R. system (Bicevskis et al., 1966). Flowsheet for Recovery of Uranium and Thorium
Feed Solution. T h e composition of the feed solution in any solvent extraction process is extremely important, and adjustments are usually made to concentration and free acid content
Table I. Fuel Inlet and Outlet Composition afler 120 Days' Cooling for a 200-MWe H.T.G.C.R. with Recycle Inlet composition. Fissile: fertile: moderator = 1:23.5 :1380
(atom ratios) Burnup = 1.0 fissions/initial fissile atom Inlet Compn., Outlet Compn., Component Kg./Core Kg./Core 78,400 78,400 Be0 13,480 14,099 ThOz 0.205 0 2aTJ02 517 497 ~8*U02 74.7 0 z34Uoz is i 55 0 2wo2 1.082 0 2 a w 0 * 2.424 101.9 239 PUO 2.86 22.3 2QPu02 11.12 6.58 "'PUOZ 0 12.81 2 4 2 ~ ~ 0 ~ 8.76 x lo' curies Fission product activity ...
178
l&EC PROCESS DESIGN A N D DEVELOPMENT
td minimize the recovery problems. I n this instance a high feed concentration was selected, since this assists in achieving high recoveries and leads to minimum waste volumes. However, the choice of free acid content in the feed was based on a compromise between two opposing requirements : T h e desirability of near-stoichiometric salt solution of low free acidity, which results in high extraction coefficients and avoids third-phase problems in TBP systems. The desirability of a highly acid solution to ensure that hydrolysis does not occur, especially in the extraction section. The concentration of the feed solution was limited by the major component, beryllium ; the highest beryllium concentration which was stable in the system was 2.5M and this was selected. A free acidity of 0.5 to 1.ON was chosen, since 0.5N appeared to be the lowest acidity that would be acceptable without introducing hydrolysis problems. Split Contactor Concept. Because of the high salting strength of the feed solution, conventional extraction-scrub flowsheets cannot be used without producing third-phase problems in a solvent-extraction system employing TBP and the diluent O M S ("odorless mineral spirits"), With the split contactor system, the scrub soldtion raffinate is not combined with the feed solution a t the feed point, but is rejected directly ; therefore, high acid concentrations d o not build u p a t the feed point. T h e unscrubbed extract is fed to a n intermediate point in the scrub section, whose upper part is used to provide decontamination from fission products, while a second solvent stream reduces product losses in the scrub raffinate to an acceptable level in the lower part. T h e actinide elements are thus largely separated from beryllium in the extraction section of the flowsheet, while major decontamination from fission products is obtained in the scrub section, where the solvent stream also ensures that losses of actinides are controlled. Using this basic concept, together with extrapolated equilibrium data, simple calculations based on the behavior of uranium, thorium, and nitric acid were used to determine stage requirements in the solvent-extraction processes (Baillie and Ryan, 1965). These resulted in a calculated flowsheet which served as a starting point for the experimental work. The split contactor concept and the calculated data are illustrated in Figure 2. Experimental Work
Preliminary testing of the flowsheet was performed in a glove box which tontained two 16-stage Westlake mixersettlers. Later, experimental work was conducted in a shielded aqueous micro processing (AMP) plant (Cairns et al., 1965), which permitted the handling of activity levels in the feed of up to 5 curies per liter. T h e plant consisted of two shielded and
SCRUBBED EXTRACT
SCRUB
1-
1
WASTE TBP
1
0.53911
1
Stager 5 Theoretical Stages SOLVENT
I
R
SCRUB RAFFINATE
n
Ratio
0 4
OOOOlM
1
FEE3
1
UNSCRUBBED EXTRACT Ratio 0 8
1: 1
PRODUCT
Theoretical Stage
I SOLVENT1
Ratio 1 0 - 5
"0, 0 0 3 1 N
BERYLLIUM RAFFINATE
0.124 g l l
Ratlo 2.5M IO
HNO, O , O N 000008M 0 0 0 0 2 /I
SPLIT CONTACTOR FLOWSHEET *OMS:ODOURLESS MINERAL SPIRITS
Figure 2.
Stream concentrations from flowsheet design calculations
contained cells, one solvent clean-up cell, and five sh'elded process tanks, all contained in a ventilated and filtered enclosure. I n all experimental operations and facilities, special attention was given to personnel safety in the handling of both radioactivity and beryllium (Stokinger, 1966). T h e experimental program was undertaken to test the calculations and determine whether the calculated flowsheet requirements in each part of the solvent-extraction process were adequate to provide the required product recoveries and decontamination factors. It was necessary also to determine the composition of the scrub solution which should be used, and to establish that the feed solution chosen was suitable. A series of runs was undertaken, initially with feed solutions containing no plutonium or radioactive fission products, later with the addition of mixed fission products, and finally with both fission products and plutonium present. I n all instances, the feed solutions were simulated by mixing the various components to give the desired feed solution concentrations. Experimental runs occupied from 10 to 96 hours, with samples of end streams being taken, in general, every hour; chemical equilibrium was reached in 6 to 8 hours of smooth operation.
b-
Results All the experimental work was done with feed solutions having a beryllium concentration of 2.5M. Free acidity in the feed solution was varied from 0.3N to 1.5N, and two concentrations of TBP, 25 and 40 volume 7 0 , were used. Initial work, without activity present, showed that losses of thorium, from the extract contactor to the beryllium raffinate stream, were dependent on the free acid concentration in the raffinate within the range examined. Losses were higher a t low free acidities, particularly when the free acid fell below 0.5N in the raffinate, as shown in Figure 3. At free acid levels below 0.3N in the feed, the mixer-settler became inoperable because of the formation of a precipitate in the extraction stages. Under these conditions the raffinate stream was acid-deficient and a free acid content of 1N in the feed was selected as the minimum required for satisfactory operation. The use of 25% TBP in the diluent was shown to be pref-
4 E X T R A C T I O N STAGES USED
VOL 7
NO. 2
APRIL 1 9 6 0
179
erable to 40% TBP in early work. With 25% the resultant extract contained less free acid, making stripping of the actinides less difficult. Also, a higher percentage actinide loading of the solvent could be achieved, enabling higher decontamination factors to be achieved. However, 25% TBP involves the use of additional extraction stages for the flow ratios given in Figure 2. The actual numbers of stages used with 25% TBP in the experimental work were: extraction section, 4; lower scrub section, 5 ; upper scrub section, 7 ; and strip section, 16. T h e first three sections were provided in the one mixer-settler, while a second mixer-settler was used for the stripping section. Theoretically only four stages are needed for stripping but it was convenient to use 16 because a 16-stage unit was installed in the AMP plant. T h e scrub contactor consists of two parts, the upper part providing decontamination and the lower part controlling the losses, Using solutions with no activity present, the free acid content of the feed to the scrub section and the composition and concentration of the scrub solution were investigated, since these affect the salting behavior of the scrub solution. Initially, it was expected that all of the salting strength necessary in the scrub section would be provided by the nitric acid in the unscrubbed extract and that the scrub solution would contain neither salt nor acid (see Figure 2). However, experimental work showed that: The nitric acid contained in the unscrubbed extract did not give rise to a n adequate salting strength in the scrub section, resulting in a n excessive loss of thorium to the scrub raffinate. The introduction of additional nitric acid in the scrub solution did not give a n acceptable flowsheet, since under the conditions necessary to provide sufficient salting strength to control losses, too much nitric acid remained in the scrubbed extract and this led to difficulties in the stripping of uranium and thorium. The necessary salting strength was therefore made up by using aluminum nitrate in the scrub solution. Conditions in the extraction section were fixed, to given an unscrubbed extract with a free acid content of 0.3 to 0.4iV,which produced a freeacid level of about 0.7N in the scrub raffinate, and less than 0.05N in the scrubbed extract, thus eliminating stripping problems. T h e concentration of aluminum nitrate in the scrub solution was then varied to determine thorium losses. Trace levels of mixed fission products were used also to investigate the effect on decontamination factors. Figure 4 indicates that optimum scrubbing occurs around 0.1 to 0.5M Al(NO3)3. Later work showed that when using a scrub solution containing 0.37M Al(NO3)a thorium losses to the scrub raffinate could be held to an acceptable 1% with uranium losses about O.Olyc. Gross gamma decontamination factors, using 0.37M Al(NO3)3, were 1 X lo3 to 5 X lo3with feed solutions containing up to 1 curie of mixed fission products per liter. N o difficulties were encountered in obtaining high recoveries in the strip section, provided that the free acid content of the extract was kept sufficiently low to ensure that the acidity in the aqueous phase was less than 0.1Nand preferably about 0.05N. Under such circumstances, the losses of both uranium and thorium could be kept well below O.1ycwhen using a water strip solution. Figure 5 gives the results of a final series of runs completed in the AMP plant. I n the runs which produced these data, we obtained mass balances accounting for materials within i5% based on final composite analyses. T h e uranium and thorium losses are given in Table 11, which shows that the original goals 180
l&EC PROCESS DESIGN A N D DEVELOPMENT
T H O R I U M 'le ,RECOVERY
I
00
I 0
THORIUM LOSSES BECOME SIGNIFICANT B E L O W A B O U T 0.3 M
01 0 2 0 3 0 4 O S 0.6 0.7 0 8 0 . 9 ALUMlNlUM NITRATE M O L A R I T Y (M)
1.0
I
1.1
Figure 4. Variation of gross Y D F T ~and thorium recovery with aluminum nitrate molarity in scrub solution
of 99% recovery of thorium and 99.8% recovery of uranium were attained. Decontamination factors for individual nuclides were measured in these experiments, rather than over-all decontamination factors. These are given in Table I11 for the major gamma emitters in the feed solution. An examination of gamma spectrograph scans of the activity in the various streams showed that significant decontamination from *03Ru, 137Cs, and 144Ce was achieved in the extraction section itself, while decontamination from 95Zr-Nb was largely obtained ?n the scrub section. In several experimental runs, trace quantities of plutonium were added to the feed solution to determine its behavior in the system. An addition of 0.05M ferrous sulfamate to both the feed solution and scrub solution was made to maintain the plutonium in the trivalent state, since plutonium rejection to
Table II.
Uranium and Thorium Losses
Per Cent Loss to Specified Stream Component
Be-containing raffinate
U Th
0.02 0.16
Table 111.
Fission Product g6Zr-Nb lO3Ru
Scrub raffinate
Stripped solvent
Total
0.01
0.02 0.02
0.05 0.96
0.78
Decontamination Factors
Over-All Dtcontamination Factor 1,100
1 3 7 c s
5,700 >1,000
ln4Ce
>1,000
SCRUB Flow r a t e 0 81 ml /.in. A I (NO,), 0 57M
S
stoqes
Fii”’
z3 e
vie
,
I : c, I
...., -
76 4 ‘lo
7De8in. water
SCRUB RAFFINATE
o ei
1
81
0 37M 0 4911 0 ooozq 0 OZqll o 7n
HNO,
OMS
1647;
0 005gl1 00003 q / I 0 OOOlq I I
Th
V
Be
o 005n
HNO, Activity- * ZrlNb95
4 a IO’
pa233
4 1 IO‘
2 I 10’
z a IO’
Pa233
UNSCRUBBED
I
16 S t o q e s
FEE0
L
*
SOLVENT (1B P Flow r a t e
TB P 0 M.S
I.30.I/
23 e
I) .in
76 4 “lo
’
--
NOTES Activity i8 given In counts per minute per nilltlitre
A
8. HMO,
C
Activrtv ZrlNb95 RulO3 Ce144 til37 Pa233
4 stages
T
I
*
Figure 5.
226gll IN 4 I I I
I IO’
I IO’ x IOb
x IO,“ 3x10
.
Th U
29 2 q / l l3qtl
Be “0,
0 0004qll
Activity-
The composition of the unscrubbed extract and scrubbed extract w r e not measured
*
C005N
2 10’ RulO3 z a 10’ Po233 I a 10’ cs Too low t o detect
Zrlnb95
Summary of solvent-extraction runs
the raffinate streams was desired. T h e results showed that the plutonium distribution in the various end streams was: beryllium raffinate, 88%; scrub raffinate, 9%; recycled organic, 0.1%; product, 2.9%. T h e rejection of approximately 90% of the plutonium in the system to the beryllium raffinate stream was satisfactory. Higher rejections could be achieved. Conclusions
Adequate product recovery and decontamination from fission products have been achieved in a TBP extraction process developed for recovering thorium and uranium from highly salted beryllium nitrate feed solution at low levels of activity. Acknowledgment
The authors acknowledge the assistance of the Analytical Chemistry Section and of the technical staff, particularly V. K. Vilkaitis, during the experimental work.
literature Cited
Baillie, M. G., Ryan, R. K., “Development of Solvent Extraction Processes for the H.T.G.C.R. Fuel Cycle,” Part 1, “Design of a Flowsheet for the Recovery of Actinides,” Australian AEC Rept. E 139 (June 1965). Bicevskis, A., Hesse, E. W., Mercer, D. J., “Thorium Fuel Cycle for a Beryllium Oxide Pebble-Bed Reactor,” Second International Thorium Fuel Cycle Symposium, Gatlinburg, Tenn., May 3-6,1966. Cairns, R. C., Baillie, M. G., Farrell, M. S., May, J. R., “Fuel Cycle Studies for a Beryllium Oxide Moderated High Temperature Gas-Cooled Reactor,” Second International Thorium Fuel Cycle Symposium, Gatlinburg, Tenn., May 3-6, 1966. Cairns, R. C., Fox, B. J., Baillie, M. G., “Small Pilot Plant for the Development of Radioactive Solvent Extraction Processes,” Institution of Engineers, Australia, Mech. Chem. Eng. Trans. 511, No. 2 , 153-60 (November 1965). Smith, R., “Development of Beryllia-Based Dispersion Fuels for the Australian H.T.G.C. Pebble-Bed Reactor Study,’’ Nuclear Metallurgy Symposium on High Temperature Nuclear Fuels, Delavan, Wis., Oct. 3-5, 1966. Stokinger, H. E., ed., “Beryllium: Its Industrial Hygiene Aspects,” Academic Press, New York, 1966. RECEIVED for review February 20, 1967 ACCEPTEDJuly 31, 1967
VOL. 7
NO.
2
APRIL
1968
181
