Ind. Eng. Chem. Res. 1999, 38, 1683-1689
1683
Flowsheet Feasibility Studies Using ABEC Resins for Removal of Pertechnetate from Nuclear Wastes Andrew H. Bond,*,†,‡ Michael J. Gula,† James T. Harvey,† Jonathan M. Duffey,† E. Philip Horwitz,*,† Scott T. Griffin,§ Robin D. Rogers,*,§ and Jack L. Collins| Eichrom Industries, Inc., 8205 S. Cass Avenue, Ste. 107, Darien, Illinois 60561, Department of Chemistry, The University of Alabama, Tuscaloosa, Alabama 35487, and Oak Ridge National Laboratory, P.O. Box 2008, MS 6223, Oak Ridge, Tennessee 37831
A flowsheet for the extraction and immobilization of 99TcO4- from radioactive wastes has been developed. The three-stage flowsheet comprises selective sorption of 99TcO4- by ABEC resins, secondary concentration of 99TcO4- from H2O by a nonselective silica-based anion-exchange resin, and encapsulation and immobilization of the radionuclide-loaded anion-exchange resin in hydrous titanium oxide microspheres. Each process has been independently tested, and the results are discussed. The utility of this flowsheet in the extraction and immobilization of 99TcO4- from alkaline radioactive wastes, neutralized nuclear fuel reprocessing streams, alkaline vitrification off-gas scrubber solutions, and solvent wash stages from the treatment of acidic high-level wastes is also presented. Introduction Treatment of radioactive wastes generated from nuclear power applications and defense processing is a global concern requiring immediate and prudent action. Radioactive wastes resulting from defense activities differ considerably from those originating from power production, with the former often varying widely in composition and the latter generally taking the form of clad fuel elements, cooling pond waters, and smaller waste volumes from purification, decontamination, and analytical activities.1-3 For wastes resulting from power production, the well-known PUREX process enables the treatment and reuse of nuclear materials for this purpose.1,4-6 The complexity of defense wastes arises from the variety of matrixes in which the 90Sr, 99Tc, 129I, 137Cs, 238U, 237Np, 239/240Pu, and 241Am reside. For example, most of the high-level wastes in the former Soviet Union are in an acidic medium and treatment by SREX,7-10 cobalt dicarbollide,10-14 TRUEX,6,9,10,15,16 or Russian TRU10,14,17-19 processes is feasible. By contrast, the acidic streams from the separation and purification of plutonium by the BiPO4, REDOX, and PUREX processes at Hanford in the United States were made alkaline to prevent corrosion of the mild steel storage tanks.20 Following the Hanford model, the Savannah River site currently stores wastes from PUREX raffinate streams that were made alkaline.21 The results of such a drastic pH swing on these wastes are the formation of sludges, supernates, and saltcakes, which contain an average activity of 1 Ci/L in the ≈2.3 × 108 L of waste at the Hanford site.22 Only a small number of constituents in these wastes are radioactive. Thus, there are significant economic incentives to * To whom correspondence should be addressed. † Eichrom Industries, Inc. ‡ Current address: Chemistry Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439. Phone: 630-252-0957. Fax: 630-252-7501. E-mail:
[email protected]. anl.gov. § The University of Alabama. | Oak Ridge National Laboratory.
segregate the radioactive constituents from the inactive components, which simplifies vitrification processing and conserves valuable repository space.22 Given the complexity of alkaline radioactive wastes, the treatment options must remain simple and must accommodate vitrification: the immobilization strategy of choice for high-level wastes.3 A variety of treatment strategies have been proposed that range from no pretreatment (i.e., bulk vitrification22) to extensive chemical separations (i.e., the CLEAN option22,23). Bulk vitrification is the most costly approach, as all radioactive and nonradioactive elements will be destined for geologic deposition22 and significant adjustments to the melter feed stream will be required to maintain the amorphous features of the borosilicate glass. Extensive chemical separations should be a less costly approach because a far smaller volume of waste will require deep geologic deposition; however, chemical and economic hurdles may exist as improvements to chemical processing infrastructure may be required and acid dissolution of intractable sludges may be complicated. Most of the actinides and 90Sr are present in the sludge layers20 because of the highly alkaline conditions. Using this existing segregation will reduce somewhat the amount of waste to be vitrified, but the supernates and dissolved saltcake solutions contain 137Cs+ and 99TcO4- that must be concentrated to minimize the final waste volume. The most feasible approach involves bulk vitrification of the sludge material and use of mobile chromatographic columns to remove the soluble 137Cs+ and 99TcO - ions. Technetium-99 poses unique problems as 4 it has a 2.14 × 105 year half-life,24 TcO4- is environmentally mobile,25,26 and Tc2O7 is volatile.25,27-29 Volatility of the heptaoxide is important in bulk vitrification because 99Tc must be scrubbed from off-gases, which generates a secondary liquid waste stream. Two options for the removal of 99TcO4- from alkaline wastes have been identified: Reillex-HPQ anionexchange resins30-34 and the ABEC resins33-40 described in the preceding paper in this issue. Both materials are capable of sorbing 99TcO4- from alkaline supernate solutions with minimal pretreatment, but Reillex-HPQ
10.1021/ie980611o CCC: $18.00 © 1999 American Chemical Society Published on Web 02/09/1999
1684 Ind. Eng. Chem. Res., Vol. 38, No. 4, 1999
is regenerated by elution with NaOH/Sn(II)/ethylenediamine or strong acid whereas the ABEC resins are stripped using H2O. This paper describes flowsheet feasibility studies using ABEC resins for the removal of 99TcO4- from alkaline radioactive wastes, its concentration on a silica-based anion-exchange resin, and ultimate immobilization of the technetium in hydrous titanium oxide (HTiO) spheres. Extension of this flowsheet to 99TcO4- removal from neutralized nuclear fuel reprocessing streams, alkaline vitrification off-gas scrubber solutions, and solvent wash stages from the treatment of high-level radioactive wastes is also presented. Experimental Section Chemicals used in the metal ion uptake studies were of ACS reagent-grade or better. All water was deionized using commercial deionization systems. Technetium-99 was assayed by liquid scintillation counting and technetium-95m by γ counting. Standard radiometric procedures were employed throughout.35-38 Chromatographic Experiments. The chromatogram in which 99TcO4- was loaded from 4.0 M NaOH was collected using a modified formulation of the geltype resin ABEC G2. This material was prepared in a 110 L reactor using Me-PEG-2000, 2% divinylbenzene (DVB), and 10% vinylbenzyl chloride (VBC) in diglyme. The resin has a mesh range of 30-60, 53.00% solids, and a bed density of 0.382 g of dry resin/mL of bed in H2O. A total of 2.1328 g of the modified air-dried ABEC G2 resin was slurried in H2O and quantitatively transferred to an 8 mm (i.d.) glass column. A small piece of glass wool was placed on top of the bed to prevent its disruption during the addition of eluent. The geometrically determined bed volume (BV) was 3.143 mL (2.959 mL calculated) in H2O and 2.829 mL in 4.0 M NaOH, with the difference corresponding to 11.10% swelling. Sodium is not retained by ABEC resins, and breakthrough of 22Na+ afforded a free column volume (FCV) of 1.538 mL in 4.0 M NaOH. The resin volume under loading conditions was calculated to be 1.291 mL from the BV and FCV data. The column was conditioned by eluting 10.0 mL (3.53 BV) of 4.0 M NaOH, followed by elution of a 4.0 M NaOH solution spiked with 0.91 mM NH499TcO4 under gravity flow at a rate of ≈1.0 mL/ min. After 322 mL (114 BV) of load solution eluted, the column was rinsed with 94.8 mL (33.5 BV) of 4.0 M NaOH and the 99TcO4- was stripped using H2O. Samples were collected in tared vials, and all volumes were calculated gravimetrically using the respective solution densities. The ABEC resin designated G2 in the preceding paper in this issue was employed for the chromatographic experiment in which a nuclear fuel reprocessing simulant was used as the load solution. ABEC G2 is a gel-type resin prepared using Me-PEG-2000, 2% DVB, and 20% VBC and has a mesh range of 50-100, 41.36% solids, and a bed density of 0.296 g of dry resin/mL of bed in H2O. The simplified simulant was prepared by dissolution of 61.88 g of Al2(SO4)3‚18H2O and 4.999 g of FeSO4‚ 7H2O in 250 mL of 2.0 M Na2SO4. After the resin was washed with H2O to remove the 10% (v/v) aqueous CH3OH shipping solution, 0.5278 g of air-dried ABEC G2 resin was slurried in H2O and quantitatively transferred to an 8 mm (i.d.) glass column. A frit was placed on top of the bed to prevent its disruption during the
addition of eluent. The BV was 0.754 mL (0.737 mL calculated) in H2O and 0.729 mL in the simulant solution, with the difference corresponding to approximately 3.43% swelling. Breakthrough of 22Na+ afforded a FCV of 0.258 mL in the simulant. The resin volume under loading conditions was calculated to be 0.471 mL from the BV and FCV data. The column was conditioned by eluting 10.0 mL (≈14 BV) of 2.0 M Na2SO4, and then the nuclear fuel reprocessing simulant, spiked with 6.99 mM NH499TcO4, was eluted under gravity flow at a rate of ≈0.67 mL/min. After 38.9 mL (53.4 BV) of load solution eluted, the column was rinsed with 9.72 mL (≈13 BV) of 3.5 M Na2SO4 and the 99TcO4- was stripped using H2O. Samples were collected and assayed as described above. Loading and Encapsulation of Anion-Exchange Resins in HTiO Microspheres. Nucleosil 100 5SB, a silica-based benzyltrimethylammonium chloride anionexchange resin of 5 µm particle diameter, was loaded with TcO4- and encapsulated in HTiO microspheres at levels of 20 and 40% (w/w). In each case, loading of TcO4- was performed first and then the resin was immobilized in the HTiO matrix. For the sample loaded to 20%, 8.557 g of H2O containing 1322 µCi of 99TcO4spiked with 95mTcO4- was contacted with 1.1610 g of Nucleosil 100 5SB (solid:solution ) 1:7.4) with gentle agitation for 48 h in a centrifuge tube. The 40% encapsulated sample was treated similarly except that 3.1037 g of Nucleosil 100 5SB was contacted with 19.25 g of H2O (solid:solution ) 1:6.2) containing 3634 µCi of 99TcO - spiked with 95mTcO -. Following contact, each 4 4 sample was centrifuged and the supernate decanted and filtered through a 0.45 µm nylon membrane filter. Analysis of the 95mTc γ emission in the filtrate was used to determine the uptake parameters reported in Table 1. HTiO microspheres containing the TcO4--loaded anionexchange resin were prepared by suspending the resin in a broth containing hexamethylenetetramine, urea, and a titanium salt. The predetermined broth formulation41 was designed to yield air-dried microspheres with compositions of 20 and 40% Nucleosil 100 5SB and 80 and 60% HTiO, respectively. After the microspheres formed and hardened in hot silicone oil at 90 °C, the solids were washed four times with trichloroethylene to remove residual oil, four times with 0.05 M NH4OH to remove unreacted starting materials, and once with H2O. Table 1 reports the activities of 99TcO4- in the silicone oil and trichloroethylene, NH4OH, and H2O washes. The microspheres were then dried at ambient temperature for 3 days. The air-dried samples were divided into three approximately equal parts, and one portion was dried at 105 °C for 5.3 h and another at 170 °C for 5.3 h. Drying the 20%-loaded microspheres at 105 °C and at 170 °C resulted in H2O losses of 17.4 and 35.9%, respectively. The 40%-loaded microspheres showed water losses of 4.7 and 22.8% at 105 and 170 °C, respectively. HTiO Microsphere Leaching Procedure. The HTiO microspheres containing the technetium-loaded anion-exchange resin were subjected to a standard Toxicity Characteristic Leaching Procedure (TCLP). The TCLP leachant solution was prepared by addition of 11.4 mL of glacial CH3CO2H to 1.0 L of H2O, followed by the addition of 128.6 mL of 1.0 M NaOH and dilution to 2.0 L. The final pH of the solution was 4.9. The microspheres were gently mixed with a specified volume
Ind. Eng. Chem. Res., Vol. 38, No. 4, 1999 1685
Figure 1. Column chromatographic extraction of 0.91 mM 99TcO - from 4.0 M NaOH by ABEC G2 (30-60 mesh, 2.83 mL 4 bed, at 25 °C). Greater than 98% of the 99TcO4- activity was removed in 10 BV of H2O.
of leachant to afford the solid:solution ratios reported in Table 2, and samples of the initial leachant solution were removed and counted after 24, 96, and 168 h. The initial leachant solutions were then decanted and the microspheres contacted with a fresh volume of leachant maintaining the same solid:solution ratios. Results and Discussion The sorption properties and capacities of ABEC resins differ as a function of the load solution composition. Extraction of TcO4- from different solutions depends largely on the aqueous phase anion and generally increases in the sequence OH- > SO42- > CO32- > PO43-.35-40 A general description of the competitive hydration uptake mechanism displayed by ABEC resins has been reported.42 ABEC Column Chromatography. Figure 1 shows the column load, rinse, and strip curves for 99TcO4using the modified 30-60 mesh ABEC G2 resin. Loading occurred from 4.0 M NaOH, where 50% breakthrough was observed after approximately 47.2 BV of the load solution was eluted. During the 33.5 BV of 4.0 M NaOH rinse, the eluate activity was only reduced to approximately one-third that of the feed activity. The H2O strip afforded 98% of the loaded 99TcO4- in 10 BV. The strip section shows a smaller trailing peak around 184 BV that is attributed to a pause in the elution. The stripping peak maximum is 17.6 times greater than the 0.91 mM concentration of 99TcO4- in the load solution, and analysis of the data yields a capacity of 19 µmol of 99TcO -/mL of bed in H O. The resin bed swelled only 4 2 11.10% during crossover to the H2O strip, which is within common equipment tolerances. Figure 2 shows the removal of 99TcO4- from a neutralized waste simulant by 50-100 mesh ABEC G2. This neutralized simulant represents a radioactive waste arising in a commercial nuclear fuel reprocessing flowsheet and contains 0.072 M FeSO4‚7H2O and 0.37 M Al2(SO4)3‚18H2O in 2.0 M Na2SO4. Fifty percent breakthrough occurred at 15.5 BV, and loading continued to 53.4 BV, where a rinse with ≈13 BV of 3.5 M Na2SO4 decreased the eluate activity 30-fold. Stripping with H2O afforded >95% of the 99TcO4- in 8 BV, which corresponds to a loading capacity of 59 µmol of 99TcO4-/ mL of bed in H2O. Only a 3.43% increase in bed height was observed during crossover from the rinse solution to the H2O strip.
Figure 2. Column chromatographic extraction of 99TcO4- from a neutralized nuclear fuel reprocessing simulant containing 6.99 mM 99TcO -, 0.072 M FeSO ‚7H O, and 0.37 M Al (SO ) ‚18H O in 2.0 4 4 2 2 4 3 2 M Na2SO4 by ABEC G2 (50-100 mesh, 0.729 mL bed, at 25 °C). 99 Greater than 95% of the TcO4 was stripped with 8 BV of H2O.
These column chromatographic experiments were designed to ensure saturation of the column with 99TcO - so that reliable capacity information could be 4 obtained. The seemingly rapid breakthrough observed in these chromatograms is an artifact of the millimolar concentrations of 99TcO4- in the feed solutions. The [99TcO4-] calculated for a Hanford Single Shell Tank (SST) waste diluted to 5 M Na+ is approximately 12 µM,23 or, put another way, the chromatograms in Figures 1 and 2 were collected using 99TcO4- concentrations of approximately 75 and 580 times, respectively, more concentrated than that of diluted SST wastes. On the basis of these chromatograms and previously published data,35-40 a decrease in the load concentration of 99TcO - to more realistic process-specific conditions and 4 the use of larger columns would most certainly yield improved concentration factors. Loading and Encapsulation of the Silica-Based Anion-Exchange Resin. Long-term disposal of 99Tccontaining low-level wastes is costly and, given that the radionuclides must be immobilized to prevent migration to the far field, every effort must be taken to minimize waste volumes and to provide chemically and radiolytically stable final waste forms. Because ABEC columns are regenerated using H2O, the 99TcO4- is eluted in a matrix well-suited not only for the conventional immobilization strategies of vitrification or grouting but also for sorption on an anion-exchange resin. Initial treatment of the radioactive wastes using anionexchange materials is not feasible because of their low selectivity for 99TcO4- over the NO2-, NO3-, and other anions present in the wastes. However, anion-exchange resins can provide an additional concentration effect after 99TcO4- is eluted from an ABEC column using H2O. Because immobilization on a material suitable for long-term storage is also an objective, silica-based anionexchange resins of 5 µm particle diameter have been investigated. Using particles of this size, 99TcO4- may be loaded onto the anion-exchange resin and further encapsulated in HTiO microspheres prior to deposition in the near surface repository. Although this encapsulation process slightly increases the volume of waste, there are considerable benefits to introducing an additional barrier to impede radionuclide mobilization. A secondary concentration and immobilization process complementing the selective removal of 99TcO4- from
1686 Ind. Eng. Chem. Res., Vol. 38, No. 4, 1999 Table 1. Activity of 99TcO4- Remaining in Solution after Contact with a Silica-Based Anion-Exchange Resin or the Resin Encapsulated in HTiO Microspheres solid
solution
µCi in solution
anion-exchange resina 20% anion-exchange resin in HTiOc 20% anion-exchange resin in HTiO 20% anion-exchange resin in HTiO 20% anion-exchange resin in HTiO 20% anion-exchange resin in HTiO
H2O H2Ob silicone oil trichloroethylene first 0.05 M NH4OH wash remaining three 0.05 M NH4OH washes H2O wash
1322 78 not detectable not detectable 42 24 2.0
anion-exchange resina 40% anion-exchange resin in HTiOc 40% anion-exchange resin in HTiO 40% anion-exchange resin in HTiO 40% anion-exchange resin in HTiO 40% anion-exchange resin in HTiO
H2O H2Od silicone oil trichloroethylene first 0.05 M NH4OH wash remaining three 0.05 M NH4OH washes H2O wash
3634 15 not detectable not detectable 91 73 7.0
% removed 94.1 3.38 1.93 0.16 99.6 2.51 2.02 0.19
a Nucleosil 100 5SB, benzyltrimethylammonium chloride resin, 5 µm particle diameter. b For a solid:solution ratio of 1:7.4 and a contact time of 48 h. c Percent loading of Nucleosil 100 5SB anion-exchange resin in HTiO microspheres. d For a solid:solution ratio of 1:6.2 and a contact time of 48 h.
Table 2. Percent Removal of
a
% resinb in HTiO
drying temp (°C)
20 20 20 40 40 40
ambient 105 170 ambient 105 170
99TcO 4
by TCLP Leachant Solutionsa as a Function of Time % 99TcO4- removed
solid:solution ratio
initial µCi 99TcO 4
24 h
leachant 1 96 h
168 h
1:338 1:463 1:457 1:389 1:352 1:458
155 141 176 285 331 313
0.18 0.07 0.19 0.16 0.16 0.22
0.17 0.07 0.19 0.17 0.16 0.21
0.28 0.25 0.19 0.25 0.23 0.22
leachant 2 24 h 96 h 0.04 0.06 0.07 0.04 0.08 0.06
0.04 0.07 0.07 0.05 0.07 0.07
Acetic acid/sodium acetate buffer at pH ) 4.9. b Nucleosil 100 5SB, benzyltrimethylammonium chloride resin, 5 µm particle diameter.
radioactive wastes by ABEC resins has been independently tested, and the results are presented in Tables 1 and 2. Sorption of 99TcO4- from H2O by the silicabased benzyltrimethylammonium anion-exchange resin known as Nucleosil 100 5SB is shown in Table 1 at two different 99TcO4- activities. For initial solution activities of 1322 and 3634 µCi, the uptake percentages are 94.1 and 99.6%, which correspond to weight distribution ratios of 120 and 1400 mL/g, respectively, after 48 h of contact. The difference in these weight distribution ratios is not understood at present, because the solid: solution ratios (1:7.4 at 1322 µCi and 1:6.2 at 3634 µCi of 99TcO4-), contact times, and method of agitation were comparable. More investigations are needed to explain this difference. Desorption of 99TcO4- from the HTiO microspheres and contamination of the processing fluids had been a potential concern. Table 1 lists the activities and percentages of 99TcO4- in the silicone oil curing medium and the trichloroethylene, 0.05 M NH4OH, and H2O washes. Desorption of 99TcO4- in the microsphere processing solvents (i.e., silicone oil and trichloroethylene) is negligible and indicates that these solvents can be reused. Some 99TcO4- was removed during microsphere washing: the initial 0.05 M NH4OH wash removed 42 and 91 µCi of 99TcO4- at the 1322 and 3634 µCi loading levels, respectively. Activities in the remaining three 0.05 M NH4OH washes then decreased to 24 and 73 µCi. Crossover to a H2O wash liberated only 2.0 and 7.0 µCi for the low- and high-activity samples, respectively. Any removal of 99TcO4- from the HTiO microspheres during these washing procedures is a shortcoming, and a proprietary process in which the 99TcO4- is sorbed onto premanufactured HTiO microspheres containing the
silica-based anion-exchange resin is being investigated.43 Such a process would also eliminate release of 99TcO - to the HTiO microsphere processing fluids. 4 That the 99TcO4- is effectively immobilized in the anion-exchange resin encapsulated HTiO microspheres is reflected in Table 2, in which the TCLP leaching results are reported. For microspheres prepared at 20 and 40% loading of the anion-exchange resin and dried at different temperatures, the percent removal of 99TcO4in the first leachant does not exceed 0.28% or 0.434 µCi from a 0.7377 g microsphere sample initially loaded with 155 µCi of 99TcO4-. After 96 h of exposure, e0.08% more 99TcO4- was removed by the second leachant solution (40%-loaded microspheres dried at 105 °C initially containing 331 µCi of 99TcO4-). The activity mobilized by the leachant does not vary systematically as a function of the microsphere drying temperature; consequently, it is not clear if drying the HTiO microspheres beyond 105 °C imparts increased stability to leaching. Comparable amounts of 99TcO4- were leached from the 20 and 40% resin-loaded microspheres despite the fact that the latter were loaded with approximately 52% more activity. To conserve space in the repository, preparation of the 40%-loaded microspheres and drying to 105 °C would be preferred. In addition to the resistance to leaching imparted by the HTiO encapsulation process, the matrix also contributes to the radiation shielding. Safely assuming the near surface repository has a minimum of three barriers to migration (i.e., concrete vaults or reservoir liner, 55-gal drum, and the HTiO microsphere surface), the 99TcO4- should be effectively immobilized and shielded over the anticipated operating lifetime of the facility. Flowsheet for the Treatment of Radioactive Wastes. The selective removal of 99TcO4- from radioac-
Ind. Eng. Chem. Res., Vol. 38, No. 4, 1999 1687
Figure 3. Example flowsheet for the removal of radioactive wastes using ABEC resins.
99TcO 4
from
tive wastes by ABEC columns, its secondary concentration on a silica-based anion-exchange resin, and immobilization in HTiO microspheres have been combined into the abbreviated flowsheet shown in Figure 3. A clarified alkaline supernate (neat, blended, or diluted to ≈4-5 M Na+), dissolved saltcake feed solution, or neutralized nuclear fuel reprocessing stream could proceed directly to 99TcO4- removal by ABEC columns because 137Cs+, if present, does not interfere with solute retention by ABEC materials. However, as shown in Figure 3, removal of the highly radioactive 137Cs+ by inorganic ion-exchange materials44 or the resorcinol formaldehyde resins44 is advantageous as remote handling and shielding requirements would be greatly diminished. The effluent from 137Cs+ decontamination (stream 1) would then proceed to 99TcO4- extraction by the ABEC columns to yield an eluate classifiable as a low-level waste. To further improve the volume reduction achieved by this flowsheet, the strip eluate from the ABEC columns (stream 2) is treated with a commercially available nonselective silica-based anion-exchange resin. This step further concentrates the 99TcO4- and provides a radiolytically stable medium suitable for immobilization using grouting or the low-temperature HTiO encapsulation process described above. The aqueous stream remaining after sorption by the anion-exchange resin is recycled to strip freshly loaded ABEC columns. Depending upon the form of the anion-exchange resin (e.g., chloride, acetate, etc.), a bleed stream may be required to prevent the accumulation of anions that could adversely affect loading of 99TcO4- by the anionexchange resin. It is acknowledged that the use of ABEC resins in such a processing flowsheet will depend on a multitude of factors, but the process stream compatibility, selectivity, ease of stripping, and flexibility regarding immobilization strategies offered by ABEC resins are unparalleled advantages. ABEC resins also may be employed in the treatment of secondary wastes generated from vitrification or
evaporative concentration processing. Studies involving vitrification of nuclear fuel reprocessing simulants have shown that 2.9% of the 99Tc is lost to the off-gas as the heptaoxide after 6 h at 1050 °C.45,46 Off-gas purification systems are employed to prevent a discharge of radioactive gas to the environment, and consist of mesh filters, liquid spray scrub towers, and filtration through highefficiency particulate (HEPA) filter manifolds prior to discharge to the atmosphere.45-47 It is the liquid wastes generated during the scrubbing regime that contain 99TcO - and, thus, require further radioactive waste 4 treatment. To neutralize the acidic nitrogen oxides present in many of the off-gases, molar concentrations of NaOH have been employed in off-gas scrubbers for thermal processing.1,45,47 Quenching by an alkaline scrubber solution readily converts Tc2O7 back to TcO4-.27 Assuming comparable conditions for the bulk vitrification of the SST wastes at Hanford results in a comparable loss of technetium to the off-gas; ≈2 × 104 g of 99TcO - would be trapped in the scrubber solution after 4 vitrification of the ≈5 × 105 g of 99Tc initially present. The flowsheet in Figure 3 could easily be adapted to permit 99TcO4- removal from these secondary wastes. The clarified alkaline scrubber solution would proceed directly to 99TcO4- loading on ABEC columns, and the eluate stream would be recycled to the scrubber. The strip eluate containing the concentrated 99TcO4- in H2O could be directed to any of the concentration and/or immobilization strategies discussed above. Another potential application of ABEC resins is apparent in the treatment of acidic high-level waste streams. Such wastes are less prevalent in the United States’ defense complex (although approximately 5.6 × 106 L of acidic waste are slated for treatment at the Idaho National Engineering and Environmental Laboratory10,22,48); however, the majority of the defense wastes in the former Soviet Union reside in acidic matrixes. The TRUEX6,9,10,15,16 or Russian TRU extraction processes10,14,17-19 are being studied intensively for the treatment of these acidic wastes. Both processes partition the trivalent actinides using neutral bifunctional carbamoylmethylphosphine oxide extractants (TRUEX, octyl,phenyl-N,N-diisobutylcarbamoylmethylphosphine oxide with tri-n-butyl phosphate as phase modifier; Russian TRU, diphenyl-N,N-di-n-butylcarbamoylmethylphosphine oxide) that form alkylphosphoric acids upon radiolysis. These degradation products interfere with the actinide stripping circuits, and to prevent such complications, a NaOH or Na2CO3 solvent wash is routinely performed prior to solvent recycle.3,6,10 This wash removes the alkylphosphoric acid radiolytic degradation products from the process solvent and also strips 99TcO4- that is coextracted and carried through the extraction stages. The alkaline solvent wash is recycled, but eventually buildup of contaminants requires that it be discarded, thus generating a 99TcO4-containing secondary waste stream. Numerous reports have appeared in which uptake of 99TcO4- by ABEC resins was demonstrated from concentrated carbonate or hydroxide solutions,35,36,38-40 indicating that removal of 99TcO4- from such solvent wash raffinates is possible. Conclusions ABEC resins may be used for the selective removal of 99TcO4- from alkaline radioactive wastes, neutralized nuclear fuel reprocessing streams, alkaline off-gas scrubber solutions, and hydroxide or carbonate solvent
1688 Ind. Eng. Chem. Res., Vol. 38, No. 4, 1999
wash stages from the treatment of high-level wastes. The selective separation and concentration of 99TcO4from these solutions by ABEC resins and its ultimate elution in H2O offers unmatched flexibility in subsequent immobilization processes. The flowsheet described in this work involves a secondary concentration step where 99TcO4- is sorbed onto a silica-based anionexchange resin which is further immobilized in a HTiO matrix. Alternatively, emplacement of the silica-based anion-exchange resin in the HTiO microspheres prior to 99TcO4- loading would greatly simplify the overall process as the strip eluate from an ABEC column could be directed to a column containing the HTiO composite microspheres.43 A future report will describe the development and testing of this process. Leaching studies have shown good resistance to mobilization of 99TcO4by an acetic acid/acetate matrix at pH ) 4.9, and when combined with the standard engineering barriers of the repository, the HTiO microspheres should provide a radiolytically and chemically stable medium for longterm near surface storage. The strip eluate from ABEC columns is also compatible with the conventional immobilization strategies of grouting or vitrification, although the latter will produce secondary radioactive wastes because of volatilization of Tc2O7. Future research will be directed toward resin cycling studies, an integrated flowsheet demonstration, and more processspecific testing for the removal of 99TcO4- from off-gas scrubber solutions and solvent wash streams from highlevel waste processing. Acknowledgment This work was funded by the U.S. Department of Energy Morgantown Energy Technology Center under Contract No. DE-AC21-97MC33137. Assistance from Cara M. Tomasek is gratefully acknowledged. Literature Cited (1) Choppin, G. R.; Rydberg, J. Nuclear Chemistry: Theory and Applications; Pergamon Press: Oxford, U.K., 1980. (2) Solvent Extraction and Ion Exchange in the Nuclear Fuel Cycle; Logsdail, D. H., Mills, A. L., Eds.; Ellis Horwood, Ltd.: Chichester, U.K., 1985. (3) Wymer, R. G. Reprocessing of Nuclear Fuel. In Proceedings of the NATO Advanced Study Institute: Chemical Separation Technologies and Related Methods of Nuclear Waste Management: Application, Problems, and Research Needs; Choppin, G. R., Khankhasayev, M., Eds.; in preparation, 1998. (4) McKibben, J. M. Chemistry of the PUREX Process; DPSPU83-272-1; Savannah River Plant: Aiken, SC, 1983. (5) Science and Technology of Tributyl Phosphate. Volume I: Synthesis, Properties, Reactions, and Analysis; Schulz, W. W., Navratil, J. D., Talbot, A. E., Eds.; CRC Press: Boca Raton, FL, 1984; Vol. I. (6) Musikas, C.; Schulz, W. W. Solvent Extraction in Nuclear Science and Technology. In Principles and Practices of Solvent Extraction; Rydberg, J., Musikas, C., Choppin, G. R., Eds.; Marcel Dekker: New York, 1992. (7) Horwitz, E. P.; Dietz, M. L.; Fisher, D. E. Extraction of Strontium from Nitric Acid Solutions Using Dicyclohexano-18crown-6 and its Derivatives. Solvent Extr. Ion Exch. 1990, 8, 557. (8) Horwitz, E. P.; Dietz, M. L.; Fisher, D. E. SREX: A New Process for the Extraction and Recovery of Strontium from Acidic Nuclear Waste Streams. Solvent Extr. Ion Exch. 1991, 9, 1. (9) Horwitz, E. P.; Dietz, M. L.; Diamond, H.; Rogers, R. D.; Leonard, R. A. Combined TRU-Sr Extraction/Recovery Process. In Solvent Extraction in the Process Industries, Proceedings of ISEC’93; Logsdail, D. H., Slater, M. J., Eds.; Elsevier Applied Science: London, 1993; Vol. 3. (10) Horwitz, E. P.; Schulz, W. W. Solvent Extraction in the Treatment of Acidic High-Level Liquid Waste: Where Do We
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Received for review September 25, 1998 Revised manuscript received December 21, 1998 Accepted December 28, 1998 IE980611O
