Ronald Scott Herbst

Development and demonstration of solvent extraction processes for the separation of radionuclides from acidic radioactive waste

The presence of long-lived radionuclides presents a challenge to the management of radioactive wastes. Immobilization of these radionuclides must be accomplished prior to long-term, permanent disposal. Separation of the radionuclides from the waste solutions has the potential of significantly decreasing the costs associated with the immobilization and disposal of the radioactive waste by minimizing waste volumes. Several solvent extraction processes have been developed and demonstrated at the Idaho National Engineering and Environmental Laboratory for the separation of transuranic elements (TRUs), 90Sr, and/or 137Cs from acidic radioactive waste solutions. The Transuranic Extraction (TRUEX) and phosphine oxide (POR) processes for the separation of TRUs, the Strontium Extraction (SREX) process for the separation of 90Sr, the chlorinated cobalt dicarbollide (ChCoDiC) process for the separation of 137Cs and 90Sr, and a universal solvent extraction process for the simultaneous separation of TRUs, 90Sr, and 137Cs have all been demonstrated in centrifugal contactors using actual radioactive waste solutions. This article summarizes the most recent results of each of the flowsheet demonstrations and allows for comparison of the technologies. The successful demonstration of these solvent extraction processes indicates that they are all viable for the treatment of acidic radioactive waste solutions.

Extraction of uranium(VI) with diamides of dipicolinic acid from nitric acid solutions

Abstract Three structural isomers of diamides of dipicolinic acid (N,N′-diethyl-N,N′-ditolyl-dipicolinamide, EtTDPA) with varying position of the methyl group on the tolyl ring have been synthesized and investigated on extractability toward U(VI). The polar diluent FS-13 was used, and distribution ratios of U(VI) were studied as a function of nitric acid, ligand, and lithium nitrate concentrations. Extractability of uranium was shown to increase with increased concentration of nitrate and ligands. Infrared spectra of organic extraction phases indicate that nitric acid is coextracted as part of the neutral metal-ligand complex with U(VI) and EtTDPA through hydrogen bonding with the carbonyl group in the amide moiety.

Development and testing of a cobalt dicarbollide based solvent extraction process for the separation of cesium and strontium from acidic tank waste

A fission product solvent extraction technology for the simultaneous extraction of Cs and Sr from acidic tank waste has been developed as a collaborative effort of the Idaho National Engineering and Environmental Laboratory (INEEL) and the Khlopin Radium Institute in St. Petersburg, Russia. The process is being developed as a potential method for treating the current five million liter inventory of acidic tank waste stored at the INEEL. The fission product extraction process is based on an immiscible organic phase comprised of chlorinated cobalt dicarbollide (CCD, Cs extractant) and polyethylene glycol (PEG, Sr extractant) dissolved in a poly-fluorinated sulfone diluent. Batch contact experiments and preliminary flowsheet testing were used to define potential solvent composition and flowsheet configuration. This information was used to specify an initial flowsheet for countercurrent testing with simulated tank waste using a 3.3-cm diameter centrifugal contactor pilot plant at the INEEL. The initial extractant composition was 0.08 M CCD, 0.6 vol% PEG-400 in a phenyl trifluoromethyl sulfone (FS-13) diluent. Approximately 1.5 L of solvent was used (with continuous recycle) to treat 43 L of simulated tank waste during 75 hr of continuous operation. Greater than 99.992% of the Sr and 97.45% of the Cs were extracted from the simulated tank waste and recovered in the strip product. The matrix components Ba (>99.6%), Pb (99.8%), and Ca (10.6%) were also extracted from the simulated tank waste and recovered in the strip product. Less than 1% of the K, Na, Fe, Zr, and Mo were extracted from the tank waste simulant. Finally, none of the analytically determined waste components were observed to build up in the organic solvent.

Development and testing of SREX flowsheets for treatment of Idaho chemical processing plant sodium-bearing waste using centrifugal contactors

Abstract Laboratory experimentation has indicated that the SREX process is effective for partitioning 90Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. A baseline flowsheet has been proposed for the treatment of sodiumbearing waste (SBW) which includes extraction of strontium from liquid SBW into the SREX solvent (0.15 M 4′,4′ (5′)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar L®), a 0.01 M nitric acid strip section to back-extract components from the loaded solvent, and a 2.0 M HNO3 solvent acidification section to equilibrate the solvent with HNO3 prior to recycle to the extraction section. The flowsheet was designed to provide a decontamination factor (DF) of >103 which will reduce the 90Sr activity in the waste solution to below the NRC Class A LLW limit of 0.04 Ci 90Sr/m3. SREX flowsheet testing was performed using sixteen stages of 5.5-cm diameter centrifugal contactors. The behavior of stable Sr and other components which are potentially ex...

Integrated AMP-PAN, TRUEX, and SREX testing. I. Extended Flowsheet Testing for Separation of Surrogate Radionuclides from Simulated Acidic Tank Waste

Three unit operations for the removal of selected fission products, actinides, and Resource Conservation and Recovery Act metals (mercury and lead) were integrated successfully and tested at the Idaho National Engineering and Environmental Laboratory (INEEL) for extended run times with simulated acidic tank waste. The unit operations were ion exchange (IX) for Cs removal, followed by transuranic extraction (TRUEX) for Eu (actinide surrogate), Hg, and Re (Tc surrogate) removal, and subsequent strontium extraction (SREX) for Sr and Pb removal. Approximately 45 L of simulated acidic tank waste were processed through three IX columns, packed with composite ammonium molybdophosphate–polyacrylonitrile (AMP–PAN) sorbent for Cs removal. The IX system was operated continuously for ∼34 hr at 22 bed volumes (BV) per hour through the first two columns, each sized at 60 cm3 BV and operated to 100% breakthrough. The experimental breakthrough data were in excellent agreement with modeling predictions based on data obtained with much smaller (1.5 cm3) columns. The third column (220 cm3 BV) was used for polishing and Cs removal after breakthrough of the upstream columns. Cesium removal was >99.83% in the IX system and interference from other species was not observed. The IX effluent was processed through a TRUEX solvent extraction flowsheet to remove Eu (Am surrogate), Hg, and Re (Tc surrogate) from the simulated waste. The TRUEX flowsheet test was performed using 23 stages of 3.3 cm centrifugal contactors, operated a total of 71.3 hr, and processed ∼41 L of the IX effluent using 1.5 L of TRUEX solvent with constant solvent recycle. The TRUEX solvent was recycled through the flowsheet an estimated 17 times. Greater than 99.999% of the Eu, 96.3% of the Hg, and 56% of the Re were extracted from the simulated feed and recovered in the strip and wash streams. Over the course of the test, there was no detectable build-up of any components in the TRUEX solvent. The raffinate from the TRUEX test was processed subsequently through a SREX solvent extraction flowsheet to remove Sr, Pb, and Re (Tc surrogate) from the simulated waste. The SREX flowsheet test was performed using the same centrifugal contactors used in the TRUEX test after reconfiguration and the addition of three stages. Approximately 51 L of TRUEX raffinate was processed through the system during 77.9 hr of continuous operation with 1.5 L of SREX solvent and continuous solvent recycle. The SREX solvent was recycled through the system an estimated 45 times without measurable build-up of any components in the solvent. Approximately 99.9% of the Sr, >99.89% of the Pb, and >96.4% of the Re were extracted from the aqueous feed to the SREX flowsheet and recovered in the strip and wash sections.

Integrated AMP-PAN, TRUEX, and SREX testing. II. Flowsheet testing for separation of radionuclides from actual acidic radioactive waste

Three separation processes for the removal of selected fission products, actinides, and Resource Conservation and Recovery Act metals (mercury and lead) have been integrated successfully and tested using actual acidic radioactive waste at the Idaho National Engineering and Environmental Laboratory (INEEL). The separation processes integrated were ion exchange for 137Cs removal, followed by TRUEX solvent extraction for actinide, Hg, and 99Tc removal, and subsequent SREX solvent extraction for 90Sr and Pb removal. A flowsheet comprising these three processes is being developed at the INEEL to reduce the activity of acidic tank waste to allow disposal, after immobilization, as an NRC Class A LLW. Approximately 1350 mL of actual INEEL tank waste was first processed through an ion exchange column for selective Cs removal. The column was packed with a composite ammonium molybdophosphate–polyacrylonitrile (AMP–PAN) sorbent. The ion-exchange system was operated at 26 bed volumes per hour and was sized at a bed volume of 2 cm3. A 137Cs removal of 99.95% was obtained in the ion exchange system without notable interference from other species. The effluent from the ion-exchange (IX) system was stored and subsequently processed several weeks later through a TRUEX solvent extraction flowsheet to separate actinides, Hg, and 99Tc from the tank waste. The TRUEX flowsheet test was performed utilizing 23 stages of 2.0-cm diameter centrifugal contactors. Removal efficiencies of 99.2%, 94.7%, and 63% were obtained for total alpha, Hg, and 99Tc, respectively. Operational problems such as flooding and/or precipitate formation were not observed during the TRUEX flowsheet test. The raffinate from the TRUEX test was stored and subsequently processed several weeks later through a SREX solvent extraction flowsheet to separate 90Sr and Pb, from the tank waste. The SREX flowsheet test was performed using the same centrifugal contactors used in the TRUEX test after reconfiguration. Approximately 99.997% of the 90Sr and 98% of the Pb were extracted with the SREX flowsheet and recovered in the strip and wash sections. In addition, approximately 93% of the remaining alpha activity was extracted and recovered in the strip section. Operational problems such as flooding and/or precipitation formation were not observed during the SREX test.

Demonstration of the UNEX Process for the Simultaneous Separation of Cesium, Strontium, and the Actinides from Actual INEEL Sodium-Bearing Waste

A universal solvent extraction (UNEX) process for the simultaneous separation of cesium, strontium, and the actinides from actual radioactive acidic tank waste was demonstrated at the Idaho National Engineering and Environmental Laboratory. The waste solution used in the countercurrent flowsheet demonstration was obtained from tank WM-185. The UNEX process uses a tertiary solvent containing 0.08 M chlorinated cobalt dicarbollide, 0.5% polyethylene glycol-400 (PEG-400), and 0.02 M diphenyl-N,N-dibutylcarbamoyl phosphine oxide (Ph2Bu2CMPO) in a diluent consisting of phenyltrifluoromethyl sulfone (FS-13). The countercurrent flowsheet demonstration was performed in a shielded cell facility using 24 stages of 2-cm diameter centrifugal contactors. Removal efficiencies of 99.4%, 99.995%, and 99.96% were obtained for 137Cs, 90Sr, and total alpha, respectively. This is sufficient to reduce the activities of 137Cs, 90Sr, and actinides in the WM-185 waste to below NRC Class A LLW requirements. Flooding and/or precipitate formation were not observed during testing. Significant amounts of the Zr (87%), Ba (>99%), Pb (98.8%), Fe (8%), Ca (10%), Mo (32%), and K (28%) were also removed from the feed with the universal solvent extraction flowsheet. 99Tc, Al, Hg, and Na were essentially inextractable (<1% extracted).

Experimental Results of NWCF Run H4 Calcine Dissolution Studies Performed in FY-98 and -99

Dissolution experiments were performed on actual samples of NWCF Run H-4 radioactive calcine in fiscal years 1998 and 1999. Run H-4 is an aluminum/sodium blend calcine. Typical dissolution data indicates that between 90-95 wt% of H-4 calcine can be dissolved using 1gram of calcine per 10 mLs of 5-8M nitric acid at boiling temperature. Two liquid raffinate solutions composed of a WM-188/aluminum nitrate blend and a WM-185/aluminum nitrate blend were converted into calcine at the NWCF. Calcine made from each blend was collected and transferred to RAL for dissolution studies. The WM-188/aluminum nitrate blend calcine was dissolved with resultant solutions used as feed material for separation treatment experimentation. The WM-185/aluminum nitrate blend calcine dissolution testing was performed to determine compositional analyses of the dissolved solution and generate UDS for solid/liquid separation experiments. Analytical fusion techniques were then used to determine compositions of the solid calcine and UDS from dissolution. The results from each of these analyses were used to calculate elemental material balances around the dissolution process, validating the experimental data. This report contains all experimental data from dissolution experiments performed using both calcine blends.