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Cesium and Strontium Separation Technologies Literature Review

Description: Integral to the Advanced Fuel Cycle Initiative (AFCI) Program’s proposed closed nuclear fuel cycle, the fission products cesium and strontium in the dissolved spent nuclear fuel stream are to be separated and managed separately. A comprehensive literature survey is presented to identify cesium and strontium separation technologies that have the highest potential and to focus research and development efforts on these technologies. Removal of these high-heat-emitting fission products reduces the radiation fields in subsequent fuel cycle reprocessing streams and provides a significant short-term (100 yr) heat source reduction in the repository. This, along with separation of actinides, may provide a substantial future improvement in the amount of fuel that could be stored in a geologic repository. The survey and review of the candidate cesium and strontium separation technologies are presented herein. Because the AFCI program intends to manage cesium and strontium together, technologies that simultaneously separate both elements are of the greatest interest, relative to technologies that separate only one of the two elements.
Date: March 1, 2004
Creator: Todd, T. A.; Todd, T. A.; Law, J. D. & Herbst, R. S.
Partner: UNT Libraries Government Documents Department

Evaluation of the Strategic Value of Fully Burnt PBMR Spent Fuel - A Report to ISPO in Response to IAEA Letter Request (2004-08-30)

Description: The IAEA needs to determine the value of imposing safeguards on the spent fuel storage at the Pebble Bed Modular Reactor (PBMR) planned for construction in the Republic of South Africa. The PBMR will use hundreds of thousands of fuel elements in the shape of small spheres (6 cm in diameter). The PBMR plant design calls for the storage on site of all the spent fuel generated during the whole life of the reactor, expected to span 40 years. The spent fuel storage system is designed (or to be designed) for a functional life of 80 years. If it is determined that the spent fuel contains materials of interest to a would-be proliferant, then safeguards would have to be imposed and maintained until the spent fuel elements are processed into a form and composition that no longer requires safeguards. The problem addressed in this report is the determination of the strategic value of the spent fuel to such a would-be proliferant.
Date: May 1, 2006
Creator: Ougouag, A. M.; Gougar, H. D. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

The extraction of rare earth elements from ICPP sodium-bearing waste and dissolved zirconium calcine by CMP and TRUEX solvents

Description: The extraction of stable isotopes of Eu and Ce was investigated from simulated sodium-bearing waste (SBW) and dissolved zirconium calcine by TRUEX and CMP solvents at the Idaho Chemical Processing Plant (ICPP). Single batch contacts were carried out in order to evaluate the rare earth behavior in the extraction, scrub, strip and wash sections for the proposed flowsheets. It has been shown that these lanthanides are efficiently extracted from the sodium-bearing wastes into either solvent, are not scrubbed and are stripped from both of the extractants with dilute HEDPA. The extraction distribution coefficients for Ce and Eu are higher in the TRUEX solvent (D{sub Ce} = 11.7, D{sub Eu} = 14.9) compared with CMP (D{sub Ce} = 9.3, D{sub Eu} = 7.23) for SBW. The extraction distribution coefficients for Ce and Eu are considerably less in the TRUEX solvent (D{sub Ce}=1.13, D{sub Eu}=1.8) than in the CMP solvent (D{sub Ce}=7.4, D{sub Eu=}6.1) for dissolved zirconium calcine feeds. The lower distribution coefficients for the extraction of lanthanides in the TRUEX/dissolved zirconium calcine system can be explained by zirconium loading of the solvent. The data obtained also confirmed that Ce and Eu can be used as non-radioactive surrogates for Am in separation experiments with acidic solutions.
Date: November 1, 1995
Creator: Todd, T.A.; Glagolenko, I.Y.; Herbst, R.S. & Brewer, K.N.
Partner: UNT Libraries Government Documents Department

Evaluation and Testing of the Cells Unit Crossflow Filter on INEEL Dissolved Calcine Slurries

Description: Development of waste treatment processes for the remediation of radioactive wastes is currently under way at the Idaho Nuclear Technology and Engineering Center (INTEC), located at the Idaho National Engineering and Environmental Laboratory (INEEL). INTEC, formerly known as the Idaho Chemical Processing Plant, previously reprocessed nuclear fuel to retrieve fissionable uranium. Liquid waste raffinates resulting from reprocessing were solidified into a granular calcine material. Approximately 4,000 m3 of calcine are presently being stored in concrete encased stainless steel bins at the INTEC. Greater than 99 weight percent of the calcine is non-radioactive inert materials. By separating radioactive and non-radioactive constituents into high and low activity fractions, a significant high-activity volume reduction can be achieved. Prior to separation, calcine dissolution must be performed. However, dissolution studies have shown a small percentage of solids present after dissolution. Undissolved solids (UDS) in solution must be removed prior to downstream processes such as solvent extraction and ion exchange. Furthermore, residual UDS in solutions have the potential to carry excess radioactivity into low activity waste fractions, if not removed. Filtration experiments were conducted at the INEEL using the Cell Unit Filter (CUF) on actual dissolved H-4 calcine and dissolved Run 1027 non-radioactive pilot plant calcine. The purpose of this testing was to evaluate the removal and operational efficiency of crossflow filtration on slurries of various solids loading. The solids loadings tested were, 0.19, 2.44 (H-4) and 7.94 (1027) weight percent, respectively. A matrix of test patterns was used to determine the effects of transmembrane pressure and axial velocity on filtrate flux. Filtrate flux rates for each solids loading displayed a high dependence on transmembrane pressure, indicating that pressure filtration resistance limits filtrate flux. Filtrate flux rates for all solids loading displayed a negative dependency on axial velocity. This would suggest axial velocities tested were efficient ...
Date: August 1, 1998
Creator: Mann, N. R. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

Development of a SREX flowsheet for the separation of strontium from dissolved INEEL zirconium calcine

Description: Laboratory experimentation has indicated that the SREX process is effective for partitioning {sup 90}Sr from acidic radioactive waste solutions located at the Idaho Nuclear Technology and Engineering Center. These laboratory results were used to develop a flowsheet for countercurrent testing of the SREX process with dissolved pilot plant calcine. Testing was performed using 24 stages of 2-cm diameter centrifugal contactors which are installed in the Remote Analytical Laboratory hot cell. Dissolved Run No.64 pilot plant calcine spiked with {sup 85}Sr was used as feed solution for the testing. The flowsheet tested consisted of an extraction section (0.15 M 4{prime},4{prime}(5{prime})-di-(tert-butylcyclohexo)-18-crown-6 and 1.5 M TBP in Isopar-L.), a 1.0 M NaNO{sub 3} scrub section to remove extracted K from the SREX solvent, a 0.01 M HNO{sub 3} strip section for the removal of Sr from the SREX solvent, a 0.25 M Na2CO{sub 3} wash section to remove degradation products from the solvent, and a 0.1 M HNO{sub 3} rinse section. The behavior of {sup 85}Sr, Na, K, Al, B, Ca, Cr, Fe, Ni, and Zr was evaluated. The described flowsheet successfully extracted {sup 85}Sr from the dissolved pilot plant calcine with a removal efficiency of 99.6%. Distribution coefficients for {sup 85}Sr ranged from 3.6 to 4.5 in the extraction section. With these distribution coefficients a removal efficiency of approximately >99.99% was expected. It was determined that the lower than expected removal efficiency can be attributed to a stage efficiency of only 60% in the extraction section. Extracted K was effectively scrubbed from the SREX solvent with the 1.0 M NaNO{sub 3} resulting in only 6.4% of the K in the HLW strip product. Sodium was not extracted from the dissolved calcine by the SREX solvent; however, the use of a 1.0 M NaNO{sub 3} scrub solution resulted in a Na concentration of ...
Date: January 1, 1999
Creator: Law, J.D.; Wood, D.J. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Mercury extraction by the TRUEX process solvent: I. Kinetics, extractable species, dependence on nitric acid concentration and stoichiometry

Description: Mercury extraction from acidic aqueous solutions by the TRUEX process solvent (0.2 M CMPO, 1.4 M TBP in n-dodecane) has not extensively been examined. Research at the Idaho Chemical Processing Plant is currently in progress to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Preliminary experiments were performed involving the extraction of Hg{sup 203}, added as HgCl{sub 2}, from 0.01 to 10 M HNO{sub 3} solutions. Mercury distribution coefficients (D{sub Hg}) range between 3 and 60 from 0.01 M to 2 M HNO{sub 3}. At higher nitric acid concentrations, i.e. 5 M HNO{sub 3} or greater, D{sub Hg} significantly decreases to values less than 1. These results indicate mercury is extracted from acidic solutions {<=}{approximately}2 M HNO{sub 3} and stripped with nitric acid solutions {>=}{approximately}5 M HNO{sub 3}. Experimental results indicate the extractable species is HgCl{sub 2} from nitrate media, i.e., chloride must be present in the nitrate feed to extract mercury. Extractions from Hg(NO{sub 3}){sub 2} solutions indicated substantially reduced distribution ratios, typically D{sub Hg}< 1, for the range of nitric acid concentrations examined (0.01 to 8 M HNO{sub 3}). Extraction of mercury, as HgCl{sub 2}, by the individual components of the TRUEX solvent was also examined from 2 M HNO{sub 3}. The diluent, n-dodecane, does not measurably extract mercury. With a 1.4 M TBP/n-dodecane solvent, D{sub Hg} {approximately}3.4 compared with D{sub Hg} {approximately}7 for the TRUEX solvent. Classical slope analysis techniques were utilized to evaluate the stoichiometric coefficients of Hg extraction independently for both CMPO and TBP.
Date: December 1, 1995
Creator: Herbst, R.S.; Brewer, K.N.; Tranter, T.J. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

TRUEX partitioning from radioactive ICPP sodium bearing waste

Description: The Idaho Chemical Processing Plant (ICPP) located at the Idaho National Engineering Laboratory in Southeast Idaho is currently evaluating several treatment technologies applicable to waste streams generated over several decades of-nuclear fuel reprocessing. Liquid sodium bearing waste (SBW), generated primarily during decontamination activities, is one of the waste streams of interest. The TRansUranic EXtraction (TRUEX) process developed at Argonne National Laboratory is currently being evaluated to separate the actinides from SBW. On a mass basis, the amount of the radioactive species in SBW are low relative to inert matrix components. Thus, the advantage of separations is a dramatic decrease in resulting volumes of high activity waste (HAW) which must be dispositioned. Numerous studies conducted at the ICPP indicate the applicability of the TRUEX process has been demonstrated; however, these studies relied on a simulated SBW surrogate for the real waste. Consequently, a series of batch contacts were performed on samples of radioactive ICPP SBW taken from tank WM-185 to verify that actual waste would behave similarly to the simulated waste. The test results with SBW from tank WM-185 indicate the TRUEX solvent effectively extracts the actinides from the samples of actual waste. Gross alpha radioactivity, attributed predominantly to Pu and Am, was reduced from 3.14E+04 dps/mL to 1.46 dps/mL in three successive batch contacts with fresh TRUEX solvent. This reduction corresponds to a decontamination factor of DF = 20,000 or 99.995% removal of the gross a activity in the feed. The TRUEX solvent also extracted the matrix components Zr, Fe, and Hg to an appreciable extent (D{sub Zr} > 10, D{sub Fe} {approx} 2, D{sub Hg} {approx}6). Iron co-extracted with the actinides can be successfully scrubbed from the organic with 0.2 M HNO{sub 3}. Mercury can be selectively partitioned from the actinides with either sodium carbonate or nitric acid ({ge} ...
Date: March 1995
Creator: Herbst, R. S.; Brewer, K. N.; Tranter, T. J. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

CMPO purity tests in the TRUEX solvent using americium-241

Description: The Transuranic Extraction (TRUEX) Process was developed by E.P. Horwitz and coworkers at Argonne National Laboratory (ANL) to separate the +4, +6, and +3 actinides from acidic aqueous solutions of nuclear wastes. Octyl (phenyl)-N-N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) is the active actinide complexant used in the TRUEX solvent. CMPO is combined with tributyl phosphate (TBP) in an organic diluent, typically n-dodecane, to form the TRUEX solvent. Small quantities of impurities in the CMPO resulting from: (1) synthesis, (2) acid hydrolysis, or (3) radiolysis can result in actinide stripping problems from the solvent. The impurity, octylphenylphosphinic acid (POPPA), ia a powerful extractant at low acid concentrations which may be formed during CMPO synthesis. Consequently, commercial CMPO may contain sufficient quantities of POPPA to significantly impact the stripping of actinides from the TRUEX solvent. The purpose of these tests was to (1) determine if commercially available CMPO is sufficiently pure to alleviate actinide stripping problems from the TRUEX process and (2) to determine if solvent cleanup methods are sufficient to purify the commercially purchased CMPO. Extraction and solvent cleanup methodologies used by Horwitz and coworkers at ANL were used to determine CMPO purity with {sup 241}Am. The improvement of the americium distribution coefficient in dilute nitric acid resulting from further purifying this CMPO is not significant enough to warrant additional CMPO purifying steps. The commercially purchased CMPO is found to be acceptable to use, as received, in a full-scale TRUEX process.
Date: December 1, 1993
Creator: Brewer, K.N.; Herbst, R.S.; Tranter, T.J. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Pyrochemical separation of radioactive components from inert materials in ICPP high-level calcined waste

Description: Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m{sup 3} of high-level waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of {sup 137}Cs and {sup 90}Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details.
Date: May 1, 1995
Creator: Del Debbio, J.A.; Nelson, L.O. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Evaluation of the Hydraulic Performance and Mass Transfer Efficiency of the CSSX Process with the Optimized Solvent in a Single Stage of 5.5-Cm Diameter Centrifugal Contactor

Description: The Caustic-Side Solvent Extraction (CSSX) process has been selected for the separation of cesium from Savannah River Site high-level waste. The solvent composition used in the CSSX process was recently optimized so that the solvent is no longer supersaturated with respect to the calixarene crown ether extractant. Hydraulic performance and mass transfer efficiency testing of a single stage of 5.5-cm ORNL-designed centrifugal contactor has been performed for the CSSX process with the optimized solvent. Maximum throughputs of the 5.5-cm centrifugal contactor, as a function of contactor rotor speed, have been measured for the extraction, scrub, strip, and wash sections of the CSSX flowsheet at the baseline organic/aqueous flow ratios (O/A) of the process, as well as at O/A's 20% higher and 20% lower than the baseline. Maximum throughputs are comparable to the design throughput of the contactor, as well as with throughputs obtained previously in a 5-cm centrifugal contactor with the non-optimized CSSX solvent formulation. The 20% variation in O/A had minimal effect on contactor throughput. Additionally, mass transfer efficiencies have been determined for the extraction and strip sections of the flowsheet. Efficiencies were lower than the process goal of greater than or equal to 80%, ranging from 72 to 75% for the extraction section and from 36 to 60% in the strip section. Increasing the mixing intensity and/or the solution level in the mixing zone of the centrifugal contactor (residence time) could potentially increase efficiencies. Several methods are available to accomplish this including (1) increasing the size of the opening in the bottom of the rotor, resulting in a contactor which is partially pumping instead of fully pumping, (2) decreasing the number of vanes in the contactor, (3) increasing the vane height, or (4) adding vanes on the rotor and baffles on the housing of the contactor. The low ...
Date: September 19, 2002
Creator: Law, J.D.; Tillotson, R.D. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Evaluation and testing of HUMASORB-CS{trademark} for the removal of radionuclides from groundwater

Description: An independent experiment to demonstrate the combined removal of the radionuclides, {sup 85}Sr and {sup 137}Cs from groundwater has been conducted with the sorbent, HUMASORB-CS. Arctech, Inc. manufactures this humic acid-based sorbent material. This sorbent material is reported to have potential for remediation of contaminated groundwater present at DOE sites. The purpose of this work was to evaluate the removal efficiency and the capacity of the sorbent. Two ion-exchange columns were assembled at the Idaho Chemical Processing Plant (ICPP) to evaluate the sorbent technology. Initial {sup 137}Cs breakthrough in both columns was observed after 22.0 and 30.2 bed volumes, respectively. Strontium-85 removal was slightly more efficient than {sup 137}Cs removal. Initial {sup 85}Sr breakthrough in both columns was observed after 29.4 and 22.7 bed volumes, respectively. Calcium, which is of concern, is the major constituent within the feed solution. Calcium is attributed to loading interference in addition to other alkaline and alkaline earth metals such as stable Sr, Mg, Na, K, and Ba. Interfering ions fill exchange sites that greatly reduce the sorbents efficiency to sorb targeted ions such as radioactive Cs and Sr. Despite high concentrations of Ca in the feed solution, Ca was not sorbed by HUMASORB-CS. Results indicate HUMASORB-CS does not sorb sodium or potassium. Sodium and potassium concentrations were consistently observed at 100% breakthrough throughout the test.
Date: January 1, 1998
Creator: Mann, N.R.; Todd, T.A. & Wood, D.J.
Partner: UNT Libraries Government Documents Department

Actinide partitioning from actual Idaho chemical processing plant acidic tank waste using centrifugal contactors

Description: The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of the actinides from acidic radioactive wastes stored at the ICPP. These efforts have culminated in a recent demonstration of the TRUEX process with actual tank waste. This demonstration was performed using 24 stages of 2-cm diameter centrifugal contactors installed in a shielded hot cell at the ICPP Remote Analytical Laboratory. An overall removal efficiency of 99.97% was obtained for the actinides. As a result, the activity of the actinides was reduced from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, which is well below the U.S. NRC Class A LLW requirement of 10 nCi/g for non-TRU waste. Iron was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Mercury was also extracted by the TRUEX solvent (76%) and stripped from the solvent in the 0.25 M Na{sub 2}CO{sub 3} wash section.
Date: October 1, 1997
Creator: Law, J.D.; Brewer, K.N. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

TRUEX process applied to radioactive Idaho Chemical Processing Plant high-level waste calcine

Description: Equal volume batch contact experiments were performed with dissolved, radioactive high-level waste (HLW) calcine and the TRansUranic EXtraction (TRUEX) process solvent. Extraction, scrub, and strip distribution coefficients (D) were obtained for the transuranic (TRU) elements in order to evaluate the efficiency of the TRUEX process in treating this waste. The extraction, scrub, and strip behavior of other elements, such as chromium, zirconium, and technetium, was also observed. A TRU alpha decontamination factor of >10,000 was achieved; after three extraction batch contacts TRU alpha activity was reduced from 1,420 nCi/g to 0.02 nCi/g. Dilute nitric acid was used to scrub extracted acid, zirconium, and iron from the solvent prior to stripping. Dilute 1-hydroxyethane, 1-1, diphosphonic acid (HEDPA) was used as a gross TRU stripping reagent to recover the extracted TRUs. Data from these batch contact experiments were used to develop a counter-current flowsheet for TRU removal using the Generic TRUEX Model (GTM). Process improvements and optimizations of the flowsheet have been evaluated using a non-radioactive dissolved calcine simulant spiked with tracers to obtain additional distribution coefficient data. These data were used in the GTM to refine the flowsheet. The flowsheet was then evaluated using a counter-current 5.5 cm centrifugal contactor pilot plant with a non-radioactive dissolved calcine simulant. The experiments involving radioactive waste provided crucial data for developing a baseline TRUEX process flowsheet which can effectively separate TRU components from ICPP high-level waste.
Date: May 1, 1996
Creator: Brewer, K.N.; Herbst, R.S.; Law, J.D.; Todd, T.A. & Olson, A.L.
Partner: UNT Libraries Government Documents Department

Experimental results: Pilot plant calcine dissolution and liquid feed stability

Description: The dissolution of simulated Idaho Chemical Processing Plant pilot plant calcines, containing none of the radioactive actinides, lanthanides or fission products, was examined to evaluate the solubility of calcine matrix materials in acidic media. This study was a necessary precursor to dissolution and optimization experiments with actual radionuclide-containing calcines. The importance of temperature, nitric acid concentration, ratio of acid volume to calcine mass, and time on the amount, as a weight percentage of calcine dissolved, was evaluated. These parameters were studied for several representative pilot plant calcine types: (1) Run No. 74 Zirconia calcine; (2) Run No. 17 Zirconia/Sodium calcine; (3) Run No. 64 Zirconia/Sodium calcine; (3) Run No. 1027 Alumina calcine; and (4) Run No. 20 Alumina/Zirconia/Sodium calcine. Statistically designed experiments with the different pilot plant calcines indicated the effect of the studied process variables on the amount of calcine dissolved decreases in the order: Acid/Calcine Ratio > Temperature > HNO{sub 3} Concentration > Dissolution Time. The following conditions are suitable to achieve greater than 90 wt. % dissolution of most Zr, Al, or Na blend calcines: (1) Maximum nitric acid concentration of 5M; (2) Minimum acid/calcine ratio of 10 mL acid/1 gram calcine; (3) Minimum dissolution temperature of 90{degrees}C; and (4) Minimum dissolution time of 30 minutes. The formation of calcium sulphate (CaSO{sub 4}) precipitates was observed in certain dissolved calcine solutions during the dissolution experiments. Consequently, a study was initiated to evaluate if and under what conditions the resulting dissolved calcine solutions would be unstable with regards to precipitate formation. The results indicate that precipitate formation in the calcine solutions prepared under the above proposed dissolution conditions are not anticipated.
Date: February 1, 1995
Creator: Herbst, R.S.; Fryer, D.S.; Brewer, K.N.; Johnson, C.K. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Mass Transfer Testing of a 12.5-cm Rotor Centrifugal Contactor

Description: TRUEX mass transfer tests were performed using a single stage commercially available 12.5 cm centrifugal contactor and stable cerium (Ce) and europium (Eu). Test conditions included throughputs ranging from 2.5 to 15 Lpm and rotor speeds of 1750 and 2250 rpm. Ce and Eu extraction forward distribution coefficients ranged from 13 to 19. The first and second stage strip back distributions were 0.5 to 1.4 and .002 to .004, respectively, throughout the dynamic test conditions studied. Visual carryover of aqueous entrainment in all organic phase samples was estimated at < 0.1 % and organic carryover into all aqueous phase samples was about ten times less. Mass transfer efficiencies of = 98 % for both Ce and Eu in the extraction section were obtained over the entire range of test conditions. The first strip stage mass transfer efficiencies ranged from 75 to 93% trending higher with increasing throughput. Second stage mass transfer was greater than 99% in all cases. Increasing the rotor speed from 1750 to 2250 rpm had no significant effect on efficiency for all throughputs tested.
Date: September 1, 2008
Creator: Meikrantz, D. H.; Garn, T. G.; Law, J. D.; Mann, N. R. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

Clean-in-Place and Reliability Testing of a Commercial 12.5-cm Annular Centrifugal Contactor at the INL

Description: The renewed interest in advancing nuclear energy has spawned the research of advanced technologies for recycling nuclear fuel. A significant portion of the advanced fuel cycle includes the recovery of selected actinides by solvent extraction methods utilizing centrifugal contactors. Although the use of centrifugal contactors for solvent extraction is widely known, their operation is not without challenges. Solutions generated from spent fuel dissolution contain unknown quantities of undissolved solids. A majority of these solids will be removed via various methods of filtration. However, smaller particles are expected to carry through to downstream solvent extraction processes and equipment. In addition, solids/precipitates brought about by mechanical or chemical upsets are another potential area of concern. During processing, particulate captured in the rotor assembly by high centrifugal forces eventually forms a cake-like structure on the inner wall introducing balance problems and negatively affecting phase separations. One of the features recently developed for larger engineering scale Annular Centrifugal Contactors (ACCs) is the Clean-In-Place (CIP) capability. Engineered spray nozzles were installed into the hollow central rotor shaft in all four quadrants of the rotor assembly. This arrangement allows for a very convenient and effective method of solids removal from within the rotor assembly.
Date: September 1, 2007
Creator: Mann, N. R.; Garn, T. G.; Meikrantz, D. H.; Law, J. D. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

Clean-in-Place and Reliability Testing of a Commercial 12.5 cm Annular Centrifugal Contactor at the INL

Description: The renewed interest in advancing nuclear energy has spawned the research of advanced technologies for recycling nuclear fuel. A significant portion of the advanced fuel cycle includes the recovery of selected actinides by solvent extraction methods utilizing centrifugal contactors. Although the use of centrifugal contactors for solvent extraction is widely known, their operation is not without challenges. Solutions generated from spent fuel dissolution contain unknown quantities of undissolved solids. A majority of these solids will be removed via various methods of filtration. However, smaller particles are expected to carry through to downstream solvent extraction processes and equipment. In addition, solids/precipitates brought about by mechanical or chemical upsets are another potential area of concern. During processing, particulate captured in the rotor assembly by high centrifugal forces eventually forms a cake-like structure on the inner wall introducing balance problems and negatively affecting phase separations. One of the features recently developed for larger engineering scale Annular Centrifugal Contactors (ACCs) is the Clean-In-Place (CIP) capability. Engineered spray nozzles were installed into the hollow central rotor shaft in all four quadrants of the rotor assembly. This arrangement allows for a very convenient and effective method of solids removal from within the rotor assembly.
Date: September 1, 2007
Creator: Mann, N. R.; Garn, T. G.; Meikrantz, D. H.; Law, J. D. & Todd, T. A.
Partner: UNT Libraries Government Documents Department

Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors

Description: The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste.
Date: October 1, 1997
Creator: Law, J.D.; Brewer, K.N.; Todd, T.A. & Olson, L.G.
Partner: UNT Libraries Government Documents Department

Demonstration of the TRUEX process for partitioning of actinides from actual ICPP tank waste using centrifugal contactors in a shielded cell facility

Description: TRUEX is being evaluated at Idaho Chemical Processing Plant (ICPP) for separating actinides from acidic radioactive waste stored at ICPP; efforts have culminated in a recent demonstration with actual tank waste. A continuous countercurrent flowsheet test was successfully completed at ICPP using waste from tank WM-183. This demonstration was performed using 24 states of 2-cm dia centrifugal contactors in the shielded hot cell at the ICPP Remote Analytical Laboratory. The flowsheet had 8 extraction stages, 5 scrub stages, 6 strip stages, 3 solvent wash stages, and 2 acid rinse stages. A centrifugal contactor stage in the scrub section was not working during testing, and the scrub feed (aqueous) solution followed the solvent into the strip section, eliminating the scrub section in the flowsheet. An overall removal efficiency of 99.97% was obtained for the actinides, reducing the activity from 457 nCi/g in the feed to 0.12 nCi/g in the aqueous raffinate, well below the NRC Class A LLW requirement of 10 nCi/g for non-TRU waste.The 0.04 M HEDPA strip section back-extracted 99.9998% of the actinide from the TRUEX solvent. Removal efficiencies of >99. 90, 99.96, 99.98, >98.89, 93.3, and 89% were obtained for {sup 241}Am, {sup 238}Pu, {sup 239}Pu, {sup 235}U, {sup 238}U, and {sup 99}Tc. Fe was partially extracted by the TRUEX solvent, resulting in 23% of the Fe exiting in the strip product. Hg was also extracted by the TRUEX solvent (73%) and stripped from the solvent in the 0.25 M Na2CO3 wash section. Only 1.4% of the Hg exited with the high activity waste strip product.
Date: September 1, 1996
Creator: Law, J.D.; Brewer, K.N.; Herbst, R.S. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

TRUEX flowsheet testing for the removal of the actinides from dissolved ICPP zirconium calcine using centrifugal contactors

Description: Solid calcine is one of the wastes under evaluation for TRU removal by the TRUEX process. The calcine must first be dissolved in nitric acid prior to the removal of TRUs and fission products. Zirconium type calcine (generated from zirconium fuel reprocessing raffinates) comprises the majority of the calcine currently stored at the ICPP. The zirconium calcines average 18.3 wt% ZrO{sub 2} and are anticipated to be the most challenging to treat with regards to TRU removal because of the large zirconium content. This paper reports the results from a countercurrent flowsheet test performed with a dissolved calcine simulant in a 2-cm centrifugal contractor pilot plant. The simulant was spiked with radioactive {sup 241}Am and {sup 95}Zr to facilitate analysis and evaluate the behavior of the actinides. Flooding and precipitate formation were observed in the strip section during the flowsheet testing. It is postulated that the flooding occurred as a result of precipitate formation. The precipitate was determined to be ZrPO{sub 4} and was likely formed due the excessive amount of Zr carried into the strip section with the organic phase. Roughly 65% of the Zr in the feed was extracted. Of the extracted Zr, 15.6% reported to the strip product and 15.1% ended up in the organic effluent, indicating the strip section was ineffective at re-extracting Zr. The poor strip section performance was probably due to the precipitation and concomitant flooding problems encountered in the test, resulting in the strip section never achieving steady state operating conditions. Despite the obvious problems encountered during the test, > 99.18% of the americium was removed from the feed in the extraction section. This may be slightly lower than the anticipated 99.9% Am removal efficiency necessary to insure the < 10 nCi/g TRU content in the LLW raffinate.
Date: December 1, 1997
Creator: Herbst, R.S.; Law, J.D.; Brewer, K.N. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Demonstration of an optimized TRUEX flowsheet for partitioning of actinides from actual ICPP sodium-bearing waste using centrifugal contactors in a shielded cell facility

Description: The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) for the separation of the actinides from acidic radioactive wastes stored at the ICPP. These efforts have culminated in recent demonstrations of the TRUEX process with actual tank waste. The first demonstration was performed in 1996 using 24 stages of 2-cm diameter centrifugal contactors and waste from tank WM-183. Based on the results of this flowsheet demonstration, the flowsheet was optimized and a second flowsheet demonstration was performed. This test also was performed using 2-cm diameter centrifugal contactors and waste from tank WM-183. However, the total number of contactor stages was reduced from 24 to 20. Also, the concentration of HEDPA in the strip solution was reduced from 0.04 M to 0.01 M in order to minimize the amount of phosphate in the HLW fraction, which would be immobilized into a glass waste form. This flowsheet demonstration was performed using centrifugal contactors installed in the shielded hot cell at the ICPP Remote Analytical Laboratory. The flowsheet tested consisted of six extraction stages, four scrub stages, six strip stages, two solvent was stages, and two acid rinse stages. An overall removal efficiency of 99.79% was obtained for the actinides. As a result, the activity of the actinides was reduced from 540 nCi/g in the feed to 0.90 nCi/g in the aqueous raffinate, which is well below the NRC Class A LLW requirement of 10 nCi/g for non-TRU waste. Removal efficiencies of 99.84%, 99.97%, 99.97%, 99.85%, and 99.76% were obtained for {sup 241}Am, {sup 238}Pu, {sup 239}Pu, {sup 235}U, and {sup 238}U, respectively.
Date: January 1, 1998
Creator: Law, J.D.; Brewer, K.N.; Herbst, R.S.; Todd, T.A. & Olson, L.G.
Partner: UNT Libraries Government Documents Department

Demonstration of various ion exchange sorbents for the removal of cesium and strontium from tan groundwater

Description: Groundwater remediation efforts are currently ongoing at the Test Area North (TAN) located at the Idaho Engineering and Environmental Laboratory (INEEL). These efforts are primarily directed towards the Technical Support Facility-05 (TSF-05) for the removal of volatile organic compounds (VOC`s). The Groundwater Treatment Facility (GWTF), positioned near the TSF-05 injection well at TAN, was installed in 1994 to pump and treat the injection well groundwater. The GWTF was designed to be operated continuously at 50 gpm. Presently the GWTF operates in batch mode attributed to {sup 137}`Cs found at higher concentrations than expected. This presence of {sup 137}Cs, along with the higher than expected concentration of suspended solids revealed a need for further testing and evaluation of the GWTF process. Two independent experiments were conducted at the GWTF with actual groundwater pumped from TSF-25 and TSF-05 injection wells. One experiment used the 3M Company`s web technology to remove Cs from the actual groundwater, while the second experiment contained a duplicate set of ion exchange columns in series. Each set having two columns, one to remove Cs, the other to remove Sr from the actual groundwater. A total of 5 Cs or Sr specific sorbents were tested in the column setups. A total of eight batches of water were processed through the GWTF with the experimental setups in place, six from TSF-25 and two from TSF-05. Downtime occasionally occurred when high differential pressures prompted the high pressure shutoff switches to shut down the 3M experiment feed pump. A pre-filter change out was initiated when this occurred. Also, experimental setups were shut down for GWTF backwash and discharge operations.
Date: August 1, 1997
Creator: Garn, T.G.; Brewer, K.N.; Tillotson, R.D. & Todd, T.A.
Partner: UNT Libraries Government Documents Department

Cross flow filtration of aqueous radioactive tank wastes

Description: The Tank Focus Area (TFA) of the Department of Energy (DOE) Office of Science and Technology addresses remediation of radioactive waste currently stored in underground tanks. Baseline technologies for treatment of tank waste can be categorized into three types of solid liquid separation: (a) removal of radioactive species that have been absorbed or precipitated, (b) pretreatment, and (c) volume reduction of sludge and wash water. Solids formed from precipitation or absorption of radioactive ions require separation from the liquid phase to permit treatment of the liquid as Low Level Waste. This basic process is used for decontamination of tank waste at the Savannah River Site (SRS). Ion exchange of radioactive ions has been proposed for other tank wastes, requiring removal of insoluble solids to prevent bed fouling and downstream contamination. Additionally, volume reduction of washed sludge solids would reduce the tank space required for interim storage of High Level Wastes. The scope of this multi-site task is to evaluate the solid/liquid separations needed to permit treatment of tank wastes to accomplish these goals. Testing has emphasized cross now filtration with metal filters to pretreat tank wastes, due to tolerance of radiation and caustic.
Date: February 1, 1997
Creator: McCabe, D.J.; Reynolds, B.A.; Todd, T.A. & Wilson, J.H.
Partner: UNT Libraries Government Documents Department

Evaluation and Testing of IONSIV IE-911 for the Removal of Cesium-137 from INEEL Tank Waste and Dissolved Calcines

Description: Development of waste treatment processes for the remediation of radioactive wastes is currently underway. A number of experiments were performed at the Idaho Nuclear Technology and Environmental Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL) with the commercially available sorbent material, IONSIV IE-911, crystalline silicotitanate (CST), manufactured by UOP LLC. The purpose of this work was to evaluate the removal efficiency, sorbent capacity and selectivity of CST for removing Cs-137 from actual and simulated acidic tank waste in addition to dissolved pilot-plant calcine solutions. The scope of this work included batch contact tests performed with non-radioactive dissolved Al and Run-64 pilot plant calcines in addition to simulants representing the average composition of tank waste. Small-scale column tests were performed with actual INEEL tank WM-183 waste, tank waste simulant, dissolved Al and Run-64 pilot plant calcine solutions. Small-scale column experiments using actual WM-183 tank waste resulted in fifty-percent Cs-137 breakthrough at approximately 589 bed volumes. Small-scale column experiments using the tank waste simulant displayed fifty-percent Cs-137 breakthrough at approximately 700 bed volumes. Small-scale column experiments using dissolved Al calcine simulant displayed fifty-percent Cs-137 breakthrough at approximately 795 bed volumes. Column experiments with dissolved Run-64, pilot plant calcine did not reach fifty-percent breakthrough throughout the test.
Date: April 1, 1999
Creator: Mann, N. R.; Todd, T. A.; Brewer, K. N.; Wood, D. J.; Tranter, T. J. & Tullock, P. A.
Partner: UNT Libraries Government Documents Department