[en] 'In this research program, Oak Ridge National Laboratory (ORNL) is collaborating with Texas A and M University in the development of highly selective inorganic ion exchangers for the removal of cesium and strontium from nuclear tank-waste and from groundwater. Inorganic ion exchangers are developed and characterized at Texas A and M University; ORNL is involved in preparing the powders in engineered forms and testing the performance of the sorbents in actual nuclear waste solutions. The Texas A and M studies are divided into two main categories: (1) exchangers for tank wastes and (2) exchangers for groundwater remediation. These are subdivided into exchangers for use in acid and alkaline solutions for tank wastes and those that can be recycled for use in groundwater remediation. The exchangers will also be considered for in situ immobilization of radionuclides. The approach will involve a combination of exchanger synthesis, structural characterization, and ion exchange behavior. ORNL has developed a technique for preparing inorganic ion exchangers in the form of spherules by a gel-sphere internal gelation process. This technology, which was developed and used for making nuclear fuels, has the potential of greatly enhancing the usability of many other special inorganic materials because of the improved flow dynamics of the spherules. Also, pure inorganic spherules can be made without the use of binders. ORNL also has access to actual nuclear waste in the form of waste tank supernatant solutions for testing the capabilities of the sorbents for removing the cesium and strontium radionuclides from actual waste solutions. The ORNL collaboration will involve the preparation of the powdered ion exchangers, developed and synthesized at Texas A and M, in the form of spherules, and evaluating the performance of the exchangers in real nuclear waste solutions. Selected sorbents will be provided by Texas A and M for potential incorporation into microspheres, and the performance of the sorbents and microspheres will be examined using actual waste supernatant solutions. This collaborative program could potentially take an exchanger from concept, synthesis, structure determination, and elucidation of exchange mechanism, to engineered product and testing on real waste streams.'

[en] Experiments were conducted to evaluate the transuranium extraction process for partitioning actinides from actual dissolved high-level radioactive waste sludge. All tests were performed at ambient temperature. Time and budget constraints permitted only two experimental campaigns. Samples of sludge from Melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign, the rinsed sludge was dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 1.8 M with a nitric acid concentration of ca. 2.9 M. About 50% of the dry mass of the sludge was dissolved. In the other campaign, the sludge was neutralized with nitric acid to destroy the carbonates, then leached with ca. 2.6 M NaOH for ca. 6 h before rinsing with the mild caustic. The sludge was then dissolved in nitric acid to produce a solution containing total metal concentrations of ca. 0.6 M with a nitric acid concentration of ca. 1.7 M. About 80% of the sludge dissolved. The dissolved sludge solution form the first campaign began gelling immediately, and a visible gel layer was observed after 8 days. In the second campaign, the solution became hazy after ca. 8 days, indicating gel formation, but did not display separated gel layers after aging for 20 days. Batch liquid-liquid equilibrium tests of both the extraction and stripping operations were conducted. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th, and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%

[en] A test was conducted to evaluate this process for selectively removing actinides from Gunite tank sludge. Mixed waste sludge from Gunite tank W-6 was subjected to the ACT*DE*CON selective leaching process. (Nearly all the TRU content was attributed to Pu.) The sludge sample was first washed with 0.01M NaOH to remove excess sodium and nitrate in the interstitial liquid supernatant. The washed wet solids were treated with the ACT*DE*CON solvent (aqueous carbonate solution containing a chelating agent and an oxidant), using a ratio of 20 ml solvent per gram wet solids. Sludge and solvent were separated by centrifugation, and the ACT*DE*CON treatment was repeated twice. Analyses showed that 71% of the solids in the sludge were dissolved while 80% of the TRU-waste components dissolved. Low separation of the TRU-waste components from other components of the sludge mixture is indicated. Almost all the U and Ca were removed from the sludge. For sludges where most of the TRU content is Pu, the ACT*DE*CON process as tested is not effective in rendering the sludge a non-TRU waste. It is recommended that ACT*DE*CON be optimized for this specific application and that other processes using different chelating and oxidizing agents be tested. Also, the ACT*DE*CON process should be tested on TRU mixed waste in which most of the TRU elements are not Pu

[en] Objective was to develop sodium titanate (ST) microspheres, made by the HMTA (hexamethylenetetramine) internal gelation process, to remove radionuclides and heavy metals from waste streams at DOE sites. to determine the optimum amount of ST that can be embedded in hydrous Ti oxide (HTO) microspheres, batches of 9.2 to 23.3% ST in HTO were prepared. Crush strength of the air-dried microspheres was found to be highest. Sr was removed from simulated supernatant by all composite microspheres; 13.2% ST/HTO worked best

[en] Short communication

[en] One of the greatest challenges facing the Department of Energy (DOE) is the remediation of the 100 million gallons of high-level and low-level radioactive waste in the underground storage tanks at its Hanford, Savannah River, Oak Ridge, Idaho, and Fernald sites. Bench-scale batch tests have been conducted with sludge from the Melton Valley Storage Tank (MVST) Facility at Oak Ridge National Laboratory (ORNL) to evaluate separation processes for use in a comprehensive sludge-processing flow sheet for concentrating the radionuclides and reducing the volumes of storage tanks wastes for final disposal. This report discusses the hot cell apparatus, the characterization of the sludge, and the results obtained from a variety of basic and acidic leaching tests of samples of sludge. Approximately 5 L of sludge/supernate from MVST W-25 was retrieved and transferred to a stainless steel tank for mixing and storage in a hot cell. Samples were centrifuged to separate the sludge liquid and the sludge solids. Air-dried samples of sludge were analyzed to determine the concentrations of radionuclides, other metals, and anions. Based upon the air-dried weight, about 41% of the centrifuged, wet sludge solids was water. The major alpha-, gamma-, and beta-emitting radionuclides in the centrifuged, wet sludge solids were ¹³⁷Cs, ⁶⁰Co, ¹⁵⁴Eu, ²⁴¹Am, ²⁴⁴Cm, ⁹⁰Sr, Pu, U, and Th. The other major metals (in addition to the U and Th) and the anions were Na, Ca, Al, K, Mg, NO₃^-, CO₃^2-, OH^-, and O_2-. The organic carbon content was 3.0 ± 1.0%. The pH was 13

[en] Bench-scale batch equilibration tests have been conducted with supernatants from two underground tanks at the Melton Valley Storage Tank (MVST) Facility at Oak Ridge National Laboratory (ORNL) to determine the effectiveness of selected ion exchangers in removing cesium, strontium, and technetium. Seven sorbents were evaluated for cesium removal, nine for strontium removal, and four for technetium removal. The results indicate that granular potassium cobalt hexacyanoferrate was the most effective of the exchangers evaluated for removing cesium from the supernatants. The powdered forms of sodium titanate (NaTiO) and cystalline silicotitanate (CST) were superior in removing the strontium; however, for the sorbents of suitable particle size for column use, titanium monohydrogen phosphate (TiHP φ), sodium titanate/polyacrylonitrile (NaTiO-PAN), and titanium monohydrogen phosphate/polyacrylonitrile (TiP-PAN) gave the best results and were about equally effective. Reillex trademark 402 was the most effective exchanger in removing the technetium; however, it was only slightly more satisfactory than Reillex trademark HPQ

[en] Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of ∼ 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for ∼ 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of ∼ 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted

[en] Bench-scale leaching tests were conducted with samples of tank waste sludge from the Melton Valley Storage Tank (MVST) Facility at Oak Ridge National Laboratory (ORNL) to evaluate separation technology processes for use in concentrating the radionuclides and reducing the volume of waste for final disposal. This paper discusses the hot cell apparatus, the characterization of the sludge, the leaching methodology, and the results obtained from a variety of basic and acidic leaching tests of samples of sludge at ambient temperature. Basic leaching tests were also conducted at 75 and 95 deg C. The major alpha-,gamma., and beta-emitting radionuclides in the centrifuged, wet sludge solids were ¹³⁷Cs, ⁶⁰Co, ¹⁵⁴Eu, ²⁴¹Am, ²⁴⁴Cm ⁹⁰Sr, Pu, U, and Th. The other major metals (in addition to the U and Th) and anions were Na, Ca, Al, K, Mg, NO₃^-,CO₃^2-, OH^-, and O^2- organic carbon content was 3.0 +/- 1.0%. The pH was 13. A surprising result was that about 93% of the ¹³⁷Cs in the centrifuged, wet sludge solids was bound in the solids and could not be solubilized by basic leaching at ambient temperature and 75 deg C. However, the solubility of the ¹³⁷Cs was enhanced by heating the sludge to 95 deg C. In one of the tests,about 42% of the ¹³⁷Cs was removed by leaching with 6.3 M NaOH at 95 deg C.Removing ¹³⁷Cs from the W-25 sludge with nitric acid was a slow process. About 13% of the ¹³⁷Cs was removed in 16 h with 3.0 M HNO₃. Only 22% of the ¹³⁷Cs was removed in 117 h usi 6.0 M HNO₃. Successive leaching of sludge solids with 0.5 M, 3.0 M, 3.0 M; and 6.0 M HNO₃ for a total mixing time of 558 h removed 84% of the ¹³⁷Cs. The use of caustic leaching prior to HNO₃ leaching, and the use of HF with HNO₃ in acidic leaching, increased the rate of ¹³⁷Cs dissolution. Gel formation proved to be one of the biggest problems associated with HNO₃ leaching of the W-25 sludge

[en] 'To expand the authors efforts to provide families of inorganic ion exchangers useful on a global scale. In carrying out this objective, they will synthesize a variety of ion exchange materials, determine their structures and where necessary alter these structures to build in the desired properties. The underlying thermodynamic, kinetic and molecular basis of ion exchange behavior will be elucidated and their suitability for nuclear waste remediation will be assessed. As of September 1, 1996, they have synthesized a number of highly selective inorganic ion exchangers, determined their crystal structures and elucidated the mechanism of exchange for a number of these exchangers.'