WSRC-TR-2001-00072

Screening Evaluation of Alternate Sorbents and Methods for Strontium
and Actinide Removal from Alkaline Salt Solution

D. T. Hobbs, M. S. Blume, and H. L. Thacker
Westinghouse Savannah River Site
Aiken, SC 29808

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1.0 Summary

The authors conducted screening strontium and actinides removal tests with crystalline silicotitanate (CST), SrTreatÒ and sodium nonatitanate (ST) samples and by precipitation upon addition of nonradioactive strontium, calcium and sodium permanganate. Several of the alternate sorbents and the precipitation method exhibited strontium and actinide removal characteristics as good as or better than the reference MST material. We recommend continued evaluation of the ST and SrTreatÒ materials and the precipitation method as alternatives to MST for strontium and actinide removal.

Keywords: Salt Disposition, Plutonium, Uranium, Neptunium, HLW

The Savannah River Site (SRS) selected an amorphous sodium titanate material, referred to as monosodium titanate to remove strontium from supernatant high-level wastes as part of the In-Tank Precipitation (ITP) process.¹ This material is of a class of hydrous metal oxides originally developed by R. Dosch and coworkers at the Sandia National Laboratory in the 1970s. Synthesis of these amorphous materials uses a sol-gel process that yields solids with a high surface area. Testing also indicated that these materials remove a number of other species from alkaline solution including the actinides.^2,3

Kilpatrick and Hobbs of the Savannah River Technology Center (SRTC) modified the synthesis of the MST, which provided a narrower particle size distribution and improvements in filtration and settling characteristics.⁴ Personnel provided this information to a number of vendors to prepare commercial quantities of the MST for the ITP operation. Two vendors, Optima Chemical Company and Boulder Scientific, successfully prepared materials that met purchase specifications for strontium removal, particle size distribution, solids concentration and alcohol content.

Actinide removal characteristics of the MST came under increasing scrutiny in the early 1990s to ensure that the MST would not sorb sufficient fissile isotopes from the waste to pose a nuclear criticality hazard.^5,6 Research also investigated whether the decontaminated liquid waste from the ITP process would meet Z-Area limits for total alpha activity.⁷ SRTC researchers conducted a number of tests to support these concerns. Results indicated that the MST effectively removes uranium and plutonium, but will not load sufficient quantities of fissile isotopes to pose a criticality concern.^8,9 None of this testing investigated the kinetics of the adsorption process.

The Salt Disposition Systems Engineering Team identified the adsorption kinetics of actinides and strontium onto MST as a technical risk for several of the processing alternatives selected for additional evaluation. They requested that the Savannah River Technology Center examine the adsorption kinetics of MST for several process alternatives.¹⁰ The first studies examined the extent and rate of adsorption of strontium, uranium, neptunium and plutonium as a function of temperature, monosodium titanate concentration, and the concentrations of sodium and adsorbing species (Sr, Pu, Np and U). Additionally, comparison tests in the design of the experiments assessed the effects of mixing, sludge solids and the presence of sodium tetraphenylborate solids. Preliminary¹¹ and final¹² reports documented findings of this testing. Analysis of the data indicated the need to perform additional kinetic testing with radioactive SRS tank waste and with simulants at lower ionic strength and MST concentrations.

The subsequent radioactive waste tests utilized a composite material prepared from archive samples from over twenty SRS tanks. Results indicated that the extent and rate of strontium, plutonium, neptunium and uranium removal with MST in radioactive waste agree with those previously measured with simulants.¹³ Additional tests with simulated waste solutions measured the extent and rate of strontium, plutonium, neptunium and uranium removal at 25°C in the presence of 0.2 and 0.4 g/L MST at 4.5 and 7.5 M sodium concentration.¹⁴

Flowsheet calculations indicate that the rate of actinide removal is a key variable in sizing equipment for the salt processing alternatives.¹⁵ MST and sludge mixtures exhibit reduced filtration fluxes compared to mixtures containing MST, sludge and tetraphenylborate solids.¹⁶ The low filtration fluxes result in notably larger equipment sizes for the Non-Eluatable Ion Exchange (CSTIX), Caustic Side Solvent Extraction (CSSX) and Direct Grout (DG) processes than that necessary in the Small Tank Tetraphenylborate Precipitation (STTP) process. Thus, the Salt Disposition Systems Engineering Team requested that SRTC evaluate alternate materials to MST.

This report describes results from screening tests evaluating strontium and actinide removal characteristics of three different titanium-containing sorbents, crystalline silicotitanate (CST) manufactured by UOP, SrTreat^Ò offered by Fortum Engineering, sodium nonatitanate developed by Clearfield and coworkers at Texas A&M University and offered commercially by Honeywell. We also report results from an alternate removal method, coprecipitation. These materials and alternate removal method may exhibit improved actinide removal kinetics and filtration characteristics compared to MST and thus merit testing.^17,18

UOP provided two CST samples, IE-910 (Lot #899371998000001) and IE-911 (Lot #899902081000009). IE-911 is the engineered form of the CST powder (IE-910) in which a binder is used to aggregate the individual CST powder particles into a pellet form suitable for use in a column configuration. Fortum Engineering (Finland) provided two samples of SrTreat^Ò (Lot #48 and Lot #49). Honeywell Performance Polymers and Chemicals (Morristown, NJ) provided two samples of sodium nonatitanate (ST) to SRTC for testing, which arrived identified as Lot #39287-5A and #39287-5B. Dr. A. Clearfield of Texas A&M provided two samples of sodium nonatitanate identified as Lot #RC-4-64B and #RC-4-23B.

We analyzed the samples for particle size distribution, particle morphology and elemental composition. Particle size measurements were determined by a laser light-scattering technique featuring a Microtrac Model SRA150 instrument suspending the samples in an alkaline salt solution having the same salt composition used in the subsequent strontium and actinide removal screening tests. Qualitative elemental analyses resulted from the analysis of the x-rays generated during scanning electron microscopic examination. We determined the titanium content of each sample quantitatively by Inductively Coupled Plasma Emission Spectroscopy following fusing in sodium peroxide at 675°C and dissolving the product in nitric acid.

We utilized simulated waste solutions having the same composition as previously tested to determine the performance of MST in removing strontium, plutonium, uranium and neptunium from a salt solution 5.6M in sodium.¹⁹ Table I provides the composition of the simulants. The solution used in Test Set #2 was a combination of residual quantities of the simulant used in Test Set #1 and that used in the Phase V testing.

Note that the actinide concentrations determined by the ICP-MS method have a much higher standard deviation for the Test Set #2 simulant compared to Test Set #1. Also, the average plutonium (ICP-MS) and neptunium concentrations are much higher in Test Set #2 versus that in Test Set #1. The average plutonium concentrations measured by the Pu TTA method are consistent with that expected based on the quantity of plutonium added to the simulant. Since the plutonium concentrations determined by the Pu TTA method agree well with that expected for both test sets, we conclude that the high average plutonium and neptunium concentrations for the initial samples taken from each test bottle of Test Set #2 result from analytical errors. Note that the average plutonium and neptunium concentrations for the control samples, analyzed on different days, agree well the Pu TTA method averages. The average uranium concentration in the control samples from the second test set is lower than that for the first test set. However, the uranium concentrations for all of the data sets does not statistically differ given the large standard deviations in the second test set of data.

Strontium and actinide removal testing utilized the same experimental method previously reported using the MST sorbent.¹⁹ We added approximately 0.030 grams of the ST to 75 mL of the salt solution equilibrated at 25°C. This quantity of ST provides the equivalent titanium content obtained upon addition of the current baseline material, MST, at 0.4 g/L. For the coprecipitation removal method, we added in sequence 0.4 mL of a 0.01 M strontium nitrate (Sr(NO₃)₂) solution, 0.5 mL of a 0.005M calcium nitrate (Ca(NO₃)₂) solution, 0.4 mL of a 0.01 M sodium permanganate (NaMnO₄) solution and 0.54 mL of a 30 wt % hydrogen peroxide (H₂O₂) solution. Researchers pulled samples from the test bottles 1.15, 2.67, 4.69, 25.1 and 169.5 hours after initiating the addition of the precipitating agents.

4.1 Physical and Chemical Characterization of Sorbents

Characterization of the samples included particle size, scanning electron microscopy (SEM) and determination of titanium content. Figure 1 provides a graph of the volume percent versus particle size for the sodium nonatitanate (ST) samples. The graph includes the results for a MST sample and the two previously tested samples (ST-0073A and ST-0073B).

Of the two samples supplied by Honeywell, sample ST-39287-5A exhibits a broader distribution than the ST-39287-5B sample. Both of these samples exhibit a similar distribution from about 5 microns up to about 88 microns, at which point sample 5A exhibits a distinct increase in the percentage of large particle sizes. In contrast sample 5B shows a higher percentage of particle size between 1 and 5 microns.

Samples ST-RC-4-23B (23B) and ST-RC-4-64B (64B) received from A. Clearfield at Texas A&M University exhibited no particles greater than 44 microns. Both materials exhibited similar particle size distributions. These samples have an average particle size smaller that those measured for the two Honeywell-supplied ST samples chiefly due to the absence of particles greater than 44 microns.

As expected, the Sr-Treat samples exhibited significantly different particle sized distributions (see Figure 2). Lot #48 is a sample of material developed for column operation and thus consists primarily of very large particles relative to MST and the ST materials. The reported particle size distribution of the Lot #48 does not accurately reflect the true particle size distribution since the material contains particles larger than the upper detection limit of the Microtrac instrument. Lot #49 material exhibits a particle size distribution more similar to the larger of the ST samples. The smaller particle size distribution of Lot #49 is expected since the vendor sieved Lot #48 material through a fine mesh screen to obtain a sample with a smaller particle size distribution.

Figure 1. Particle Size Distributions of Sodium Nonatitanate and MST Samples

Figure 2. Particle Size Distribution of SrTreatÒ and MST Samples

Table II provides the titanium content of each received sample. Titanium concentrations ranged from 17.2 to 41.7 wt %. We used the titanium content to determine how much of each sample to add in the sorbate removal tests.

4.2 Strontium and Actinide Removal with SrTreat, Crystalline Silicotitanate and Sodium Nonatitanate Sorbents

We evaluated strontium and actinide removal with three different sorbent materials:

SrTreat^Ò, crystalline silicotitanate (CST) and sodium nonatitanate (ST) and by a coprecipitation method from a 5.6M sodium salt solution previously used in testing with MST.¹⁹ The quantity of the titanium-containing sorbent added to the tests provided the equivalent titanium content as provided by 0.4 g/L MST. Table III provides a listing of the decontamination factors and distribution constants for strontium and actinides for the tested sorbents after contacting for 24 hours and 7 days.

Table III. Decontamination Factors and Distribution Constants (K_d) for Tested Sorbents

Figure 3 (SrTreat^Ò and CST samples) and Figure 4 (ST samples) provide graphs of the strontium concentration versus time for each of the tested sorbents. Each graph includes the results for a sample of a MST reference sample tested at the same time as the other materials.

SrTreat^Ò sample Lot #49 exhibited similar strontium removal capacity to MST but with slower kinetics. SrTreat^Ò sample Lot #48 exhibited much lower strontium removal capacity than MST and approximately an order of magnitude lower removal capacity than the Lot #49 sample.

Figure 3. Strontium Removal with SrTreat^Ò, Crystalline Silicotitanate and Monosodium Titanate Samples

Figure 4. Strontium Removal with Sodium Nonatitanate and Monosodium Titanate Samples

The CST samples exhibited much poorer strontium removal characteristics than MST. The form of the CST did not significantly affect the 24-hour and 7-day capacity, but did affect initial removal kinetics. Immediate removal of strontium occurs with the IE-910, which is the powdered form of CST. In contrast, strontium removal initiates after about 4 hours of contact with the IE-911, which is the engineered form of CST.

The ST samples exhibited good strontium removal capacities and rates. Sample ST-RC-4-64B exhibited removal characteristics equal to or greater than MST. Samples ST-RC-4-23B and ST-39287-5A exhibited similar characteristics to each other with slightly lower removal capacities and slower rates as compared to MST. Sample ST-29387-5B exhibited the lowest removal capacity and slowest rate of the four samples.

Current estimates of ⁹⁰Sr decontamination factors (DFs) range from 4.5 for the average waste concentration to 26 for the bounding waste concentration.¹⁵ These factors are based on waste solution at 6.0M in sodium. Based on the measured 24-hour DFs for these eight sorbents (see Table III), five of the samples exhibited DFs higher than the required values to meet Z-Area limits for ⁹⁰Sr at both the average and bounding waste concentrations. The four samples include SrTreat^Ò Lot #49 and all four of the ST samples, ST-RC-4-64B, ST-RC-4-23B, ST-39287-5A and ST-39287-5B. SrTreat^Ò Lot #48 does not achieve the needed DF for either concentration. Both of the CST samples exhibited DFs for average waste concentration that would meet the Z-Area limit , but not for the bounding waste concentration.

4.2.2 Plutonium Removal

Testing results indicated that the SrTreat^Ò and CST materials removed approximately a factor of ten less plutonium than the reference MST sample at the equivalent quantity of added titanium (see Table III and Figure 5). The SrTreat^Ò Lot #49 and CST IE-911 samples exhibited similar plutonium removal characteristics, which are improved compared to the other SrTreat^Ò sample and the powder form of CST (IE-910). The plutonium removal trends for the CST materials are opposite those observed with strontium. Previously, McCabe reported low plutonium removal (ca. 21%) with the powder form of CST upon contact with radioactive waste solution from SRS Tank 43H.²⁰ The absence of measurable removal in this test with the powdered CST may reflect the much higher concentrations of the actinides in the simulant used in this testing compared to that in the Tank 43H waste solution. The higher plutonium removal in the engineered form of the CST compared to the powdered CST suggests that the binder present in the engineered form of CST (IE-911) may be involved in plutonium removal.

The ST samples exhibited much better plutonium removal characteristics than the SrTreat^Ò and CST samples (see Table III and Figure 6). Samples ST-RC-4-64B, ST-39287-5A and ST-39287-5B exhibited plutonium removal characteristics comparable to or exceeding that for the reference MST sample. We attribute the improved performance of these samples from Texas A&M and Honeywell versus earlier samples supplied by Honeywell to a combination of improvements in the ST structure and the addition of sufficient sorbent based on titanium content.

Figure 6. Plutonium Removal with Sodium Nonatitanate and Monosodium Titanate Samples

The plutonium decontamination factors (DFs) required to meet the Z-Area limit range from 12 for the average waste concentration to 55 for the bounding waste concentration.¹⁵ None of the sorbent samples (including MST) exhibited a DF at the bounding waste case and only one sample (ST-39287-5B) exhibited a DF meeting the average waste case after 24 hours of contact. Thus, research efforts for improving sorbent properties need to focus on increased plutonium capacity.

Results indicated similar trends for uranium removal as those observed for plutonium with the three sorbent materials (see Table III and Figures 7 and 8). We did not detect the removal of uranium with any of the SrTreat^Ò and CST samples. The ST samples removed between 33 and 67% of the uranium, which is comparable to that removed by the reference MST sample. Given the much higher uranium concentration and limited quantity of sorbent, DFs for uranium are much lower than those obtained with plutonium.

4.2.4 Neptunium Removal

Testing results indicated that the SrTreat^Ò and CST samples exhibited lower neptunium removal than the reference MST sample (see Table III and Figure 9). Compared to MST, the removal rates for these materials are slower than with MST. SrTreat^Ò Lot #49 and CST IE-911 samples exhibited greater neptunium removal than the SrTreat^Ò Lot #48 and CST IE-910 samples.

McCabe reported greater than 80% removal of the neptunium in a sample of supernatant liquid from Tank 43H upon contact with the powder form of.²⁰ The absence of measurable removal in this test with the powdered form of CST may reflect the much higher concentrations of the actinides in the simulant used in this testing compared to that in the Tank 43H waste solution. As with plutonium, the higher neptunium removal in the engineered form of the CST compared to the powdered CST suggests that the binder present in the engineered form of CST (IE-911) may be involved in neptunium removal.

The ST samples generally exhibited greater neptunium removal than the SrTreat^Ò and CST samples and comparable removal to that of the reference MST sample at the equivalent titanium addition (see Table III and Figure 10). Neptunium removal rates appeared similar for the ST samples compared to those observed for MST. Given the larger particle size distribution of the ST samples compared to MST, neptunium removal kinetics appears controlled by transport from the sorbent surface to the sorption site.

Neptunium removal is not required in the average waste case, but is in the bounding waste case (DF = 33).¹⁵ After 24 hours of contact, sorbents exhibiting better neptunium removal had DFs ranging from about 3.7 to 5.6, which is well below a DF of 33. The neptunium concentration in the simulant used in these tests is at the bounding case (approximately 450 mg/L). Thus, meeting Z-Area limits for neptunium removal at the bounding case may be at risk if increased neptunium removal cannot be achieved.

Figure 9. Neptunium Removal with SrTreat^Ò , Crystalline Silicotitanate and Monosodium Titanate Samples

4.3 Strontium and Actinide Removal by Precipitation

We performed a single test in duplicate at 25°C evaluating the removal of strontium, plutonium, uranium and neptunium upon the sequential addition at 15-minute intervals of nonradioactive strontium nitrate, calcium nitrate and sodium permanganate. Beginning two minutes after the addition of the sodium permanganate we made five separate 135-mL additions of 30 wt % hydrogen peroxide over an eleven-minute time period. The hydrogen peroxide reduces the permanganate to manganese (IV), which then precipitates as a hydrous manganese oxide, MnO₂^.xH₂O.

Table IV provides the average DFs for strontium and the three actinides obtained in this test along with the DFs measured with the reference MST material and the best ST sample. Figures 11 – 14 provide graphs of concentration versus time for each of the sorbates respectively.

Results indicate that this method featured very rapid removal of strontium and the actinides. Maximum removal occurred within one hour of the initial strontium addition, which was about 5 minutes after the completion of the sodium peroxide addition. As evidenced by higher DFs after 24 hours, the precipitation method removed more of each sorbate at the test conditions than 0.4 g/L MST.

With the exception of strontium, the 7-day results indicated lower DFs compared to MST. The decrease in DF in the precipitation test indicates dissolution of a portion of the precipitate sorbate. This behavior may indicate that the hydrogen peroxide directly reduced the actinide to a lower oxidation state with a lower solubility, which with time and contact with dissolved oxygen or other oxidant (e.g., nitrate or nitrite), reoxidized and dissolved into solution.

Figure 11. Strontium Removal upon Addition of Sr²⁺/Ca²⁺/MnO₂

Figure 13. Uranium Removal upon Addition Sr²⁺/Ca²⁺/MnO₂

The authors conducted screening strontium and actinides removal tests with crystalline silicotitanate, SrTreat^Ò and sodium nonatitanate (ST) samples and by precipitation upon addition of nonradioactive strontium, calcium and sodium permanganate. Several of the alternate sorbents (i.e., SrTreat^Ò and ST) and the precipitation method exhibited strontium and actinide removal characteristics as good as or better than the reference MST material. We recommend continued evaluation of the ST and SrTreat^Ò materials and the precipitation method as alternatives to MST for a batch process to remove strontium and actinides. Efforts need to focus on improving plutonium and neptunium removal characteristics to ensure that Z-Area limits can be met at bounding waste conditions.

This work used the following task plan.

This document provides deliverables for the screening evaluation of sodium nonatitanate and other materials and methods requested in the authorizing task request,

Notebooks WSRC-NB-2000-00063 and WSRC-NB-2000-00120 (D. T. Hobbs) contain the experimental data obtained from this work.

The authors thank A. Clearfield of Texas A&M University and S. Yates of Honeywell Performance Polymers and Chemicals for supplying the sodium nonatitanate samples, UOP for supplying the crystalline silicotitanate samples, P. Augustyn of Graver Technologies and E. Tusko of Fortum Engineering Oy for supplying the SrTreatÒ samples, and E. A. Kyser for supplying the actinide materials used in preparing the simulated waste solutions. We also thank D. Diprete, W. Boyce, and other members of the Analytical Developmental Section of the SRTC for performing the many radiochemical analyses.

1. D. D. Walker and M. A. Schmitz, "Technical Data Summary In-Tank Precipitation Processing of Soluble High-Level Waste", Report DPSTD-84-103, Savannah River Plant, May 1984.
2. R. W. Lynch, Ed., "Sandia Solidification Process Cumulative Report", Report SAND-76-0105, Sandia National Laboratory, January 1976.
3. W. W. Schulz, J. W. Koenst and D. R. Talant, "Application of Inorganic Sorbents in Actinide Separation Processes", ACS Symposium Series 117, J. D. Navratil and W. W. Schulz, Eds., American Chemical Society, Washington, D. C., 1980, pages 17-32.
4. "Procurement Specification for Monosodium Titanate", Specification No. Z-SPP-H-00001, Rev. 2, May 1992.
5. M. C. Chandler, "Nuclear Criticality Safety Bounding Analysis for the In-Tank Precipitation (ITP) Process (U)", Report WSRC-TR-93-171, Rev. 0, Savannah River Site, March 12, 1993.
6. C. E. Bess, "Nuclear Criticality Safety Bounding Analysis for the In-Tank Precipitation (ITP) Process, Impacted by Fissile Isotopic Weight Fractions (U)", Report WSRC-TR-94-004, Rev. 0, Savannah River Site, April 22, 1994.
7. "Process Requirements 241-82H Control Room (U)", WSRC-IM-91-63, Rev. 20, October 1998.
8. D. T. Hobbs and D. D. Walker, "Plutonium and Uranium Adsorption on Monosodium Titanate (U)", Report WSRC-RP-92-93, Savannah River Site, August 13, 1992.
9. D. T. Hobbs and S. D. Fleischman, " Fissile Solubility and Monosodium Titanate Loading Tests (U)", Report WSRC-RP-92-1273, Savannah River Site, February 12, 1993.
10. P. L. Rutland, "MST Alpha Removal and Hg Removal for Salt Team Phase 3 Evaluation", HLE-TAR-98062, Rev. 0, Savannah River Site, July 15, 1998.
11. D. T. Hobbs, M. G. Bronikowski, and W. R. Wilmarth, "Preliminary Report on Monosodium Titanate Adsorption Kinetics", Report WSRC-TR-98-00347, Rev. 0, Savannah River Site, October 5, 1998.
12. D. T. Hobbs, M. G. Bronikowski, T. B. Edwards and R. L. Pulmano, "Final Report on Phase III Testing of Monosodium Titanate Adsorption Kinetics", Report WSRC-TR-99-00134, Rev. 0, Savannah River Site, May 28, 1999.
13. D. T. Hobbs and R. L. Pulmano, "Phase IV Testing of Monosodium Titanate Adsorption with Radioactive Waste", Report WSRC-TR-99-00286, Rev. 0, Savannah River Site, September 3, 1999.
14. D. T. Hobbs and R. L. Pulmano, "Phase IV Simulant Testing of Monosodium Titanate Adsorption Kinetics", Report WSRC-TR-99-00219, Savannah River Site, June 29, 1999.
15. R. A. Dimena, O. E. Duarte, H. H. Elder, J. R. Fowler, R. C. Fowler, M. V. Gregory, T. Hang, R. A. Jacobs, P. K. Paul, J. A. Pike, P. L. Rutland, F. G. Smith III, S. G. Subosits and G. A. Taylor, "Bases, Assumptions, and Results of the Flowsheet Calculations for the Short List Salt Disposition Alternatives", Report WSRC-RP-00-00006, Rev. 1, Savannah River Site, October 2000.
16. H. H. Saito, M. R. Poirier and J. L. Siler, "Effect of Sludge Solids to Mono-sodium Titanate (MST) Ratio on MST-Treated Sludge Slurry Cross-Flow Filtration Rates", Report WSRC-TR-99-00342, Rev. 0, Savannah River Site, September 15, 1999.
17. R. A. Jacobs, Technical Task Request, HLW-SDT-TTR-99-33.0, Savannah River Site, December 1999.
18. D. T. Hobbs, "Evaluation of Alternate Materials and Methods for Strontium and Alpha Removal from Savannah River Site High-Level Waste Solutions", Report WSRC-TR-2000-00229, August 2000.
19. D. T. Hobbs, M. S. Blume and H. L. Thacker, "Phase V Simulant Testing of Monosodium Titanate Adsorption Kinetics", Report WSRC-TR-2000-00142, Rev. 0, May 24, 2000.
20. D. J. McCabe, "Examination of Crystalline Silicotitanate Applicability in Removal of Cesium from SRS High Level Waste (U)", WSRC-TR-97-0016, Rev. 0, April 25, 1997.