Search NASA⌕ Search

SEARCH · Search NASA

Results for “distribution coefficient”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Ion exchange behavior of astatine and bismuth

Astatine and bismuth sorption on several ion exchange resins from nitric acid media has been studied. This work covers commercially available resins Dowex 50 × 4, MP thiol, Dowex 1 × 8, and TEVA. One of the main advantages of using ion exchangers for astatine separation and purification is the absence of any organic media in the effluent in comparison with extraction chromatography resins. The behavior of the above-mentioned metals was investigated in up to 4 M HNO 3 solutions. The determined distribution coefficients are greater than 1 for all the resins studied, reaching 10 4 for MP thiol and TEVA resins. The interaction of nitric acid with an anion exchanger in Cl-form results in an exchange reaction, where AtO + chloro complexes can form. To understand the astatine behavior under these conditions, stability constants of AtOCl and AtOCl 2 – complexes have been reevaluated and discrepancy in the literature values has been eliminated. For each resin a thermodynamic model has been developed to suggest a possible mechanism of astatine sorption. Literature and new experimental data on bismuth sorption by studying resins in nitric acid media have been reviewed and a mathematical model to describe its behavior has been suggested. Furthermore, a ratio of corresponding fit functions of At and Bi assigned to the same resin and acidity has been used to estimate the separation factors of these elements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Hanford Tank Waste Matrix Impact on Ion Exchange Performance Using Crystalline Silicotitanate

The removal of radiocesium from Hanford tank waste supernate is a critical step in preparing feed for low-activity waste immobilization. This study evaluated cesium ion exchange performance using crystalline silicotitanate (CST) media in a series of tests designed to evaluate the influence of waste matrix variability on capacity and kinetics. Tank waste supernate subsampled from five Hanford double-shell tanks encompassed a range of sodium, hydroxide, nitrate, and nitrite concentrations in order to assess the impact of feed variability on the performance of the ion exchange system. Both equilibrium and dynamic ion exchange tests were conducted to quantify cesium distribution coefficients and breakthrough behavior under prototypic operating conditions. Results indicated that effective cesium capacity varied by up to a factor of five across the matrices tested, with higher sodium concentrations significantly reducing uptake. Kinetic behavior was similarly matrix-dependent, with solution viscosity contributing to a twofold variation in mass-transfer rates. These results demonstrate the strong dependence of CST ion exchange performance on waste composition and must be incorporated into predictive models for future treatment system design and optimization.

Westesen, Amy M.↗
DOE OSTI2018

Development of Dynamic Model of a Ternary Pumped Storage Hydropower Plant

Hydropower is the largest source of renewable energy in the United States. Pumped storage hydropower (PSH) is a hydropower plant that can generate power when it operates as a hydro turbine, and it can absorb power from the electric grid when it is pumping water from the lower reservoir to the upper reservoir. The conventional role of PSH is as an energy arbitrage: it sells power when the price of electricity is high (i.e., during peak hours of energy demand), and it is operated as a pump during times when the price of electricity is low (e.g., midnight to dawn). PSH is an important part in hydropower generation. Currently, the total installed capacity is approximately 22 GW. Recently, a new type of PSH was developed, called ternary PSH (T-PSH). T-PSH is a type of advanced PSH that is different from conventional PSH and adjusted-speed PSH. T-PSH uses a single generator rotating in one direction with two different runners connected to the same shaft (i.e., one is for pumping water, and the other is for generating power as a turbine). This paper presents the development of a dynamic model of T-PSH for power system analysis. T-PSH can be operated in three different modes: generating mode, pumping mode, and hydraulic short-circuit (HSC) mode. The three modes of operation will be presented in this paper. A new combined governor is developed to represent the HSC operation mode, which is a special mode for T-PSH. The power command is divided into two parts, and a distribution coefficient is introduced to regulate the transitions from one mode to the other. Simulation and validation results are presented to verify the performance of the governor model and the characteristics of T-PSH.

15 GEOTHERMAL ENERGY↗
DOE OSTI2017

Synthesis and properties of In{sub x}Al{sub y}Ga{sub 1–x–y}P{sub z}As{sub 1–z}/GaAs heterostructures

Phase equilibria in the In–Al–Ga–P–As system have been analyzed in terms of a simple solution model. We have calculated the stability limits of InAlGaPAs solid solutions on GaAs substrates in the temperature range 937–1223 K. The addition of indium to AlGaPAs solid solutions leads to a slight decrease in growth rate under diffusion control and a considerable increase under kinetic control. Moreover, the addition of indium to AlGaPAs solid solutions has been shown to reduce the full width at half maximum of their X-ray rocking curves and photoluminescence spectra and increase the photoluminescence intensity. Increasing the indium concentration in In{sub x}Al{sub y}Ga{sub 1–x–y}P{sub z}As{sub 1–z} solid solutions to x > 0.3 extends their stability region, increases their band gap, and reduces the composition region of lattice-matched In{sub x}Al{sub y}Ga{sub 1–x–y}P{sub z}As{sub 1–z}/GaAs heterostructures, but these become thermal expansion-matched. With increasing indium concentration in the solid solutions, the distribution coefficients of P and As decrease, whereas those of Al and In increase.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Magnetic Nanofluid Rare Earth Element Extraction Process Report, Techno Economic Analysis, and Results for Geothermal Fluids

This GDR submission is an interim technical report and raw data files from the first year of testing on functionalized nanoparticles for rare earth element extraction from geothermal fluids. The report contains Rare Earth Element uptake results (percent removal, mg Rare Earth Element/gram of sorbent, distribution coefficient) for the elements of Neodymium, Europium, Yttrium, Dysprosium, and Cesium. A detailed techno economic analysis is also presented in the report for a scaled up geothermal rare earth element extraction process. All rare earth element uptake testing was done on simulated geothermal brines with one rare earth element in each brine. The rare earth element uptake testing was conducted at room temperature.

15 GEOTHERMAL ENERGY↗
DOE OSTI2018

Effect of seawater intrusion on radioactive strontium (90Sr) sorption and transport at nuclear power plants

Batch sorption and column experiments were conducted to investigate and compare sorption and transport behavior of 90Sr on the assumption of seawater intrusion at nuclear power plants. Batch sorption experiments were carried out on fractured rocks and bedrocks using synthetic groundwater and seawater both spiked with 90Sr. In general, higher 90Sr sorption occurred on fractured rock samples than on bedrocks, because of the presence of weathered secondary minerals (iron oxide and clay) on fractured rock surfaces. However, one particular bedrock sample (WSP-B) which has higher porosity and carbon amount than fractured rock samples also showed the higher 90Sr sorption than its comparable fractured rocks. For all batch sorption studies, 90Sr sorption distribution coefficient, Kd decreased from groundwater to seawater environment due to the higher ionic strength of seawater (6.4×10-1–7.7×10-1 M) compared to groundwater (4.0×10-3–6.0×10-3 M). The three different ionic strength solutions were used in column experiments, and the results showed that transport behavior of Sr through a fractured rock had similar sorption trend to batch sorption results. The highest mobility (or least retardation) for Sr was found for 100% seawater solution compared to the highest retardation (or least mobility) for 100% groundwater solution. These sorption and transport data of Sr on solid materials contacted with various ionic strength solutions corroborate empirically defensible information for assessment of radioactive contamination in groundwater below the NPP sites located nearby shores. In addition, the experimental data will be incorporated to improve transport models of 90Sr in the subsurface environment for severe nuclear accidents.

Chang, Seeun↗
DOE OSTI2022

Effects of Microstructural Composition, Porosity, and Microcracks on the Elastic Moduli of Nuclear Graphites

The elastic responses of nuclear graphite not only depend on the graphitic content itself but are largely dictated by the microstructural constitution of the material. The types of raw materials combined with the manufacturing processes used to produce the graphite yield the microstructural content that primarily includes graphite filler, graphitized pitch binder, and voids/defects that typically occupy approximately 20% of the volume. Among these microstructural components, porosity and microcracking (considered to be part of voids/defects) heavily influence the overall properties of the material, including the elastic moduli. It is widely accepted that the primary effect of oxidation is to increase porosity, but the related effect on the moduli cannot be explained satisfactorily by simply noting changes to porosity. Models describing the elastic moduli of porous, polycrystalline graphite materials have been developed to interpret experimental determinations of Young's modulus and shear modulus in oxidized graphites. Beyond porosity, the moduli are heavily influenced by microcracks, and their effects can be assessed using physical property models based on orientation distribution coefficients and microcrack-modified, single-crystal moduli to represent the combined effects of crystallite orientation and microcracking. This work demonstrates how most modulus measurements in nuclear graphites can be understood using relatively simple models that describe the effects of porosity and microcracking. We also present directions to be pursued to account for microstructure-related effects that occur as a result of neutron irradiation.

Spicer, James B.↗
DOE OSTI2016

Effect of Selected Modeling Assumptions on Subsurface Radionuclide Transport Projections for the Potential Environmental Management Disposal Facility at Oak Ridge, Tennessee

The Department of Energy’s Office of Environmental Management recently revised a Remedial Investigation/ Feasibility Study (RI/FS) that included an analysis of subsurface radionuclide transport at a potential new Environmental Management Disposal Facility (EMDF) in East Bear Creek Valley near Oak Ridge, Tennessee. The effect of three simplifying assumptions used in the RI/FS analyses are investigated using the same subsurface pathway conceptualization but with more flexible modeling tools. Neglect of vadose zone dispersion was found to be conservative or non-conservative, depending on the retarded travel time and the half-life. For a given equilibrium distribution coefficient, a relatively narrow range of half-life was identified for which neglect of vadose zone transport is non-conservative and radionuclide discharge into surface water is non-negligible. However, there are two additional conservative simplifications in the reference case that compensate for the non-conservative effect of neglecting vadose zone dispersion: the use of a steady infiltration rate and vadose zone velocity, and the way equilibrium sorption is used to represent transport in the fractured material of the saturated aquifer. With more realistic representations of all three processes, the RI/FS reference case was found to either provide a reasonably good approximation to the peak concentration or was significantly conservative (pessimistic) for all parameter combinations considered.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99 Tc desorption K d (distribution coefficient) values for 99 Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99 Tc(IV)-bearing solid phases for 99 Tc transport in reducing conditions to support the IDF PA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Radionuclide field lysimeter experiment (RadFLEx): geochemical and hydrological data for SRS performance assessments

The SRNL Radiological Field Lysimeter Experiment (RadFLEx) is a one-of-a-kind test bed facility designed to study radionuclide geochemical processes in the Savannah River Site (SRS) vadose zone at a larger spatial scale (from grams to tens of kilograms of sediment) and temporal scale (from months to decade) than is readily afforded through laboratory studies. RadFLEx is a decade-long project that was initiated on July 5, 2012 and is funded by six different sources. The objective of this status report is as follows: 1) to report findings to date that have an impact on SRS performance assessment (PA) calculations, and 2) to provide performance metrics of the RadFLEx program. The PA results are focused on measurements of transport parameters, such as distribution coefficients (Kd values), solubility, and unsaturated flow values. As this is an interim report, additional information from subsequent research may influence our interpretation of current results. Research related to basic understanding of radionuclide geochemistry in these vadose zone soils and other source terms are not described here but are referenced for the interested reader.

58 GEOSCIENCES↗
DOE OSTI2018

PORFLOW Simulations Supporting the Saltstone Performance Assessment Revision

The most recent Performance Assessment (PA) of the Z-Area Saltstone Disposal Facility (SDF) was performed in 2009, and three Special Analyses were subsequently conducted. Currently the SDF PA analysis is being revised by the Savannah River Remediation (SRR) Waste Disposal Authority group to reflect updated input information and assumptions, and improved understanding of disposal system performance. In support of the PA revision, Savannah River National Laboratory has developed revised methods for simulating the transport of I-129 and Tc-99 through porous media in a shrinking core model using the PORFLOW code. The revised methods are based on updated reduction capacity, I-129 distribution coefficient, and Tc-99 solubility values from SRR. PORFLOW simulations of aquifer transport have been revised to reflect a recent update to the General Separations Area groundwater flow model and improved treatment of plume dispersion. Hydraulic properties for some soils / sediments have been revised based on E-Area PA maintenance work. Example simulation results are provided to illustrate the behavior of the new I-129 and Tc-99 shrinking core models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Cesium Ion Exchange Using Crystalline Silicotitanate with 5.6 M Sodium Simulant

The Low-Activity Waste Pretreatment System (LAWPS) facility is planned to pretreat Hanford tank waste supernate by filtering to remove solids and processing through ion exchange columns to remove cesium. The ion exchange media is currently (since 2018) targeted to be crystalline silicotitanate (CST). This material has been produced as an engineered form (generally spherical beads) by Honeywell UOP LLC (Des Plaines, IL). Previous forms of this media have been tested in a wide array of simulants and process scales, but column breakthrough and batch contact testing to date on actual tank waste has been limited to a few tank wastes. Washington River Protection Solutions requested a series of studies to demonstrate Cs load behavior in a small column format and determine if a simple batch contact test with CST can be used to qualify tank waste supernate prior to processing it in the LAWPS. The goal of the batch contact testing is to determine if there are any issues with the tank waste that would preclude Cs removal by the ion exchanger. Testing was conducted with IONSIV TM R9140-B CST provided by Honeywell UOP in 2018 (Batch 2081000057). The CST was received in the sodium (Na) form, so no pretreatment conditioning to convert to the Na form was required. Before use in testing, the CST was passed through a #25 sieve (710-micron openings) and collected on a #60 sieve (250-micron openings) with the intent of removing any large and fine particles. Physical properties were measured on the CST including particle size distribution (d50 = 570 microns) and scanning electron micrographic examination. The CST was rinsed with water until the contact solution was visually nearly clear. The CST bed density (1.00 g/mL) and CST bed void fraction (0.656) were measured on the sieved and rinsed CST. The batch contact testing varied Cs concentration in a 5.6 M Na simple simulant solution. Testing was conducted in a liquid volume to solids mass ratio of 200. The distribution coefficient (K d ) was determined to be 770 mL/g at the Cs equilibrium condition of 8.0 μg/mL; with a bed density of 1.00 g/mL, this corresponds to a predicted 50% Cs breakthrough of 770 bed volumes (BVs). The Cs load capacity at 8.0 μg/mL equilibrium condition was determined to be 6.16 mg Cs/g dry CST. The column testing was prototypic to the intended LAWPS operations in a lead-lag column format, albeit on a small-scale basis with 10-mL CST beds. The feed was processed downflow through the lead column and then through the lag column. Three flowrates were tested: 1.19, 1.99, and 4.56 BV/h. The feed was displaced with 0.1 M NaOH, and then the columns were rinsed with water sequentially through the lead then lag columns. One column system was evaluated top to bottom with a gamma scanner to assess the axial Cs loading. Table ES.1 summarizes the relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Crystalline Silicotitanate ion exchange media long-term storage evaluation

Crystalline Silicotitanate a (CST) ion exchange media is produced in an engineered form, known as IONSIV™ IE-911. This material has been tested extensively for use in removing 137 Cs from Savannah River Site (SRS) and Hanford site tank waste. This inorganic material is highly selective for removing cesium ions, even in the presence of high concentrations of sodium and potassium ions and is tolerant of the harsh conditions of the varying tank chemistries. A batch of caustic washed IE-911 that had been produced by the supplier and received at SRS in 2001 was tested to verify that it was still in good condition after the extended storage. This testing involved standard batch contacts to measure the removal of Cs from simulated tank waste. This report describes the results of testing performed to verify that the IE-911 matches the performance of previous batches of this media. To confirm that the performance of IE-911 is similar to previous batches and that no degradation had occurred during storage, batch contact testing was performed using SRNL’s standard methodology using a non-radioactive SRS Average Waste Simulant composition utilized in previous studies. Batch contact kinetics testing was also conducted to determine cesium loading on this batch versus time. The cesium equilibrium loading isotherm in SRS simulant at 23°C was compared to ZAM isotherm model predictions and previous test results with IE-911 media. The measured and predicted distribution coefficients and cesium loading (“Q”) are shown in Table ES-1. The physical properties of the material were also determined, including density, moisture content, and particle size.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Integrated characterization of CO 2 storage reservoirs on the Rock Springs Uplift combining geomechanics, geochemistry, and flow modeling

This project improves understanding of the effects CO 2 injection and storage on geomechanical, petrophysical, and other properties on two potential reservoirs, the Madison Limestone and Weber Sandstone, at the Rock Springs Uplift, southwest Wyoming. Rock physics studies, laboratory experiments, and computer simulations were employed. Two groups of rock samples were prepared, one aged with formation brine and the second aged with CO 2 -saturated formation brine and subsequently used for geomechanical, geochemical, and petrophysical testing. Rock physics studies provided a facies classification to discriminate among different lithologies and to investigate the feasibility of predicting petrophysical properties using seismic inversion. By applying a zone-wise regression relation established in this facies analysis, a 3D reservoir model was developed. It was determined that time-lapse seismic data are not feasible for monitoring CO 2 displacement at the Rock Springs Uplift or to mitigate leakage-risks situations because pressure and fluid saturation have small effects on elastic velocities and synthetic seismograms. The small variations that do occur are due to the low porosity of the sandstone and the high stiffness of the dolomite. In the Madison Limestone, mineral dissolution due to reaction with CO 2 -saturated brine was observed in highly permeable pathways while mineral precipitation happened in intergranular or micropores. Changes in pore size were quantified via Time-Domain Nuclear Magnetic Resonance transverse relaxation time (T 2 ) and diffusion coefficient distributions. As opposed to continuous injection of CO 2 -saturated brine, a subtle increase in porosity after static ‘soaking’ with CO 2 -saturated brine increases permeability due to the heterogeneity of changes in pore and pore throat sizes. Porosity-permeability relationships are fitted using a power-law function that can be adopted to model time-evolved porosity and permeability in a transport-limited CO 2 injection process. The effect of CO 2 -saturated brine on the linear elastic properties of Weber Sandstone and Madison Limestone cannot be generally concluded. Considering horizontal and vertical orientations and different confining pressures, no consistent relationship between stress (or strain) data and reaction with CO 2 -saturated brine was observed in the nonlinear plastic and post-failure regime. The change of elastic constants due to reaction with CO 2 -saturated brine is more significant in Madison Limestone than in Weber Sandstone; however, no consistent trend was observed. Reservoir simulations suggest that injectivity for Weber Sandstone is adversely impacted by low permeability. Changes in confining pressure do not impact injection rate and can be ignored while the effect of changes to pore pressure cannot. The effect of confining pressure and pore pressure on the size of the CO 2 plume in the Madison Limestone is negligible because of higher permeability. Geomechanical and petrophysical properties induced by CO 2 -brine-rock reactions were time-dependent. Simulations considered these time-dependent properties and indicate that injectivity, porosity distribution, pressure distribution and CO 2 saturation were affected.

58 GEOSCIENCES↗
DOE OSTI2019

Characterization and CST Batch Contact Equilibrium Testing of Aged TCCR Tank 10H Batch 1A and 2 Process Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Previous equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that CST cesium removal performance was lower than predicted by ZAM modeling for TCCR Tank 10H Production Supernate Batch 1A. An average cesium distribution coefficient of 1,948 mL/g which corresponds to 94.1% Cs + removal was observed for Batch 1A supernate at 38 °C in January of 2019 using laboratory pretreated IONSIV™ R9120-B a CST from Lot #209900034 (predominant TCCR CST batch) at a phase ratio of 122 mL/g. This cesium loading was approximately 32% lower than model predictions, presumably due to ion exchange competition or fouling associated with alkaline earth or transition metals present in the solutions at low levels. The current testing was designed to: 1) evaluate whether supernate compositional changes during storage would result in improved cesium removal performance with CST, and 2) assist in media selection and handling protocols for future TCCR columns by evaluating several CST manufacturer batches as well as samples of one CST batch which had been pretreated using two different methods (field and laboratory methods).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Characterization and CST Batch Contact Equilibrium Testing of Modified Tank 9H Process Supernate Samples in Support of TCCR

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate (CST) media to process radioactive waste solutions for the removal of ¹³⁷Cs. TCCR currently focuses on dissolving Savannah River Site (SRS) radioactive tank waste (primarily sodium saltcake solids) within Tank 10H followed by at-tank ion exchange column treatment. Two supernate batches from Tank 10H have been processed through the TCCR unit and processing of a third batch is expected soon. After processing of this supernate batch, plans are to replace the CST columns and process dissolved salt solution from Tank 9H through Tank 10H and then through the new CST columns installed in the TCCR unit. The new columns are expected to contain either a media similar to an archived CST batch (IE-911) or the R9120-B CST media used in the current TCCR columns (the two materials are fundamentally the same; just different specifications, product names, and preconditioning steps). Samples of Tank 9H dissolved saltcake were received at the Savannah River National Laboratory (SRNL) and characterized. The Tank 9H supernate contained a high sodium concentration (~9.6 M Na⁺) and will require dilution to near 6 M [Na⁺] prior to processing through the TCCR unit. The cesium concentration in Tank 9H is currently significantly higher than was observed with Tank 10H. Three dilutions of the Tank 9H supernate were conducted to mimic possible dilutions that could be conducted in the tank farm prior to TCCR processing using inhibited water, sodium hydroxide, and sodium nitrate solutions. Dilution #1 was prepared by diluting the Tank 9H supernate by a factor of 1.6 with inhibited water, Dilution #2 was prepared by diluting the Tank 9H supernate by a factor of 3.7 with a mixed sodium hydroxide/sodium nitrate diluent, and Dilution #3 was prepared by also diluting by a factor of 3.7 but with sodium hydroxide only. All three dilutions have similar Na⁺ concentrations (~6 M), but Dilutions #2 and #3 contain significantly less cesium. Major and key minor components of the diluted Tank 9H supernate samples are provided in Table ES-1. Batch contact equilibrium tests were conducted with the Tank 9H dilutions and the two different CST media batches being considered for use in the new TCCR columns. Results are summarized in Table ES-2 and compared to ZAM model predictions. The highest cesium distribution coefficient and percent removal were observed with Dilution #3. IE-911 CST (an archived CST media batch) was more effective at removing cesium than the more recently prepared R9120-B CST media, though both media samples removed >88% of the cesium and the differences may not be statistically different considering the overall uncertainty. Based on the results, maximum cesium loadings were calculated for each CST media type and Tank 9H dilution. In general, maximum cesium loadings from dilutions of this supernate batch are quite high (approaching 0.1 mmol total Cs⁺/g CST for Dilution #1). The highest calculated maximum ¹³⁷Cs loading for the Tank 9H dilutions using the ZAM model with input of the tank compositions and batch contact results was 207 Ci/kg CST (Dilution #1 with IE-911 CST). In all cases, higher maximum cesium loading values were predicted for IE-911 CST versus R9120-B, although the differences varied considerably between the dilutions. The maximum loading value for IE-911 CST with Tank 9H Dilution #1 was only 7% higher than the maximum loading for R9120-B. The maximum loading value for IE-911 CST with Dilution #3 was 24% higher than the maximum loading for R9120-B. Dilution #2 was intermediate between these values. Note that these maximum loading values are the theoretical calculated values, and actual operating conditions will cause differences. A CST binder dilution (correction) factor near 0.7 (relative to pure powder CST) was required for each batch contact test with IE-911 engineered CST using the three Tank 9H dilutions. Correction factors calculated for R9120-B CST ranged from 0.56 to 0.66 for the three dilutions. Following the batch contact tests, the CST samples were isolated from the Tank 9H solution, washed, dried, and digested in acid following established procedures. The analysis results are provided in Table ES-3. Total cesium loading values determined by CST digestion were similar to the calculated loading values based on solution analysis (Table ES-2) for all samples. In addition, the CST was observed to load calcium (R9120-B sample only), iron, strontium, lead, uranium, and plutonium after contact with the waste supernate, as has been observed previously.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

PORFLOW modeling of Vadose zone flow and transport for the E-area intermediate level vault

In support of the E-Area Performance Assessment, a two-dimensional model of water flow and radionuclide transport through the E-Area Intermediate Level Vault (ILV) and local vadose zone has been developed using the PORFLOWTM software. The purpose of the model is to calculate flux to the water table for radionuclides eluted from the ILV during its operational life, the period of institutional control, and times following site closure. Results of model calculations will be used by a three-dimensional PORFLOW model of transport through the aquifer to determine radionuclide concentrations at a hypothetical 100 meter well and at the site boundary where contaminated groundwater could be assessed by members of the public following site closure. While newly developed, the model structure closely follows that used in the 2008 PA while incorporating a refined computational mesh, updated material properties for the vadose zone soil and vault concrete, revised infiltration rates, and new solid-liquid distribution coefficients. The model also addresses degradation of concrete hydraulic properties by blending soil and concrete water retention curves over a 500 year period. That is, the hydraulic properties of ILV concrete start out as fresh concrete at the time of site closure and degrade to soil properties over a 500 year time period. This approach has not been used previously in E-Area PA’s. The model has been used to calculate water flow through the ILV vadose zone and, on a trial basis, transport of a limited number of radionuclides from the waste region inside the ILV to the water table. Figure ES-1 provides a view of the model and an example of water flow after the concrete has substantially degraded. Figure dimensions are in centimeters. The red area in the left hand side figure is the waste disposal region within the ILV. Colored bands in the figures indicate different soil and concrete regions. Figure ES-2 shows a calculation of flux to the water table (mol/yr) for U-234 and its one-year half-life decay chain (U-234 → Th-230 → Ra-226 → Pb-210). Pb-210 flux is too small to show on the same scale used for Th-230 and Ra-226 in Figure ES-2. The model has also been used to make a preliminary evaluation of the release of tritium from TPBAR disposal containers that will be place in the ILV. A separately calculated tritium release rate as a function of time was used as a source in the PORFLOW model. The TPBAR disposal schedule and tritium release rate differ from the values used in the 2008 PA. Results from these initial trial runs are compared to calculations made for the 2008 PA ILV analysis. In brief, a new PORFLOW model of flow and transport through the ILV vadose zone has been developed and tested. The model is ready to be used for PA calculations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗