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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.

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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Sulfur Solubility Testing and Characterization of LAW Phase 1 Matrix Glasses

In this report, the Savannah River National Laboratory (SRNL) provides chemical analysis results for a series of simulated low-activity waste (LAW) glass compositions. These data will be used in the development of improved sulfur solubility models for LAW glass. A procedure developed at Pacific Northwest National Laboratory (PNNL) for producing sulfur saturated melts (SSMs) was carried out at both SRNL and PNNL to fabricate the glasses characterized in this report. This method includes triplicate melting steps with excess sodium sulfate, followed by grinding and washing to remove unincorporated sulfur salts. The wash solutions were also analyzed as part of this study.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Data Validation Package - June 2015 Groundwater and Surface Water Sampling at the Green River, Utah, Disposal Site

Groundwater samples were collected during the 2015 sampling event from point-of-compliance (POC) wells 0171, 0173, 0176, 0179, 0181, and 0813 to monitor the disposition of contaminants in the middle sandstone unit of the Cedar Mountain Formation. Groundwater samples also were collected from alluvium monitoring wells 0188, 0189, 0192, 0194, and 0707, and basal sandstone monitoring wells 0182, 0184, 0185, and 0588 as a best management practice. Surface locations 0846 and 0847 were sampled to monitor for degradation of water quality in the backwater area of Brown’s Wash and in the Green River immediately downstream of Brown’s Wash. The Green River location 0801 is upstream from the site and is sampled to determine background-threshold values (BTVs). Sampling and analyses were conducted as specified in Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated, http://energy.gov/lm/downloads/sampling-and- analysis-plan-us-department-energy-office-legacy-management-sites). Water levels were measured at each sampled well. The analytical data and associated qualifiers can be viewed in environmental database reports and are also available for viewing with dynamic mapping via the GEMS (Geospatial Environmental Mapping System) website at http://gems.lm.doe.gov/#. All six POC wells are completed in the middle sandstone unit of the Cedar Mountain Formation and are monitored to measure contaminant concentrations for comparison to proposed alternate concentration limits (ACLs), as provided in Table 1. Contaminant concentrations in the POC wells remain below their respective ACLs.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

AMORE Mo-99 Spike Test Results

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solution is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Sulfur Solubility Testing and Characterization of Hanford LAW Phase 2, Inner Layer Matrix Glasses

In this report, the Savannah River National Laboratory (SRNL) provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated low activity waste (LAW) glass compositions. A procedure developed at the Pacific Northwest National Laboratory (PNNL) for producing sulfur saturated melts (SSMs) was carried out at both SRNL and PNNL to fabricate the glasses characterized in this report. This method includes triplicate melting steps with excess sodium sulfate, followed by grinding and washing to remove unincorporated sulfur salts. The wash solutions were also analyzed as part of this study. These data will be used in the development of improved sulfur solubility models for LAW glass.

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

Characterization of the LAWB99-series and ORLEC-series Glasses

In this report, the Savannah River National Laboratory provides chemical analysis results for a series of simulated low activity waste (LAW) glass compositions. These data will be used in the development of improved sulfur solubility models for LAW glass. A procedure developed at the Pacific Northwest National Laboratory for producing sulfur saturated melts (SSMs) was used to fabricate the glasses characterized in this report. This method includes triplicate melting steps with excess sodium sulfate, followed by grinding and washing to remove unincorporated sulfur salts. The wash solutions were also analyzed as part of this study.

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

Sulfur Solubility Testing and Characterization of Hanford LAW Phase 2, Outer Layer Matrix Glasses

In this report, the Savannah River National Laboratory (SRNL) provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated Low-Activity Waste (LAW) glass compositions. A procedure developed at the Pacific Northwest National Laboratory (PNNL) for producing sulfur saturated melts (SSMs) was carried out at both SRNL and PNNL to fabricate the glasses characterized in this report. This method includes triplicate melting steps with excess sodium sulfate, followed by grinding and washing to remove unincorporated sulfur salts. The wash solutions were also analyzed as part of this study. These data will be used in the development of improved property/composition models for LAW glass.

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

Sulfur Solubility Testing and Characterization of Hanford LAW Phase 2, Outer Layer Matrix Glasses

In this report, the Savannah River National Laboratory (SRNL) provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated Low-Activity Waste (LAW) glass compositions. A procedure developed at the Pacific Northwest National Laboratory (PNNL) for producing sulfur saturated melts (SSMs) was carried out at both SRNL and PNNL to fabricate the glasses characterized in this report. This method inclu des triplicate melting steps with excess sodium sulfate, followed by grinding and washing to remove unincorporated sulfur salts. The wash solutions were also analyzed as part of this study. These data will be used in the development of improved property/composition models for LAW glass.

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

Characterization of Hanford LAW Phase 3 Glasses

This report provides analyses of glass compositions, wash solution compositions, and Product Consistency Test (PCT) leachate compositions for a series of simulated low-activity waste (LAW) glasses fabricated at the Pacific Northwest National Laboratory (PNNL). These data will be used in the development of improved property/composition models for LAW glass at Hanford. Chemical analyses were performed on a representative sample of each of the quenched and sulfur saturated glasses to allow for comparisons with the targeted compositions. For some of the quenched glasses, measured concentrations of Cr 2 O 3 , K 2 O, Na 2 O, P 2 O 5 , and ZrO 2 were below the targeted values. For some of the sulfur saturated glasses, the measured concentrations of Al 2 O 3 , Cl - , Cr 2 O 3 , F - , K 2 O, Na 2 O, P 2 O 5 , V 2 O 5 , and ZrO 2 were generally low relative to the targeted values. A comparison of the measured compositions of the quenched and sulfur saturated versions of the study glasses showed that the measured concentrations of Al 2 O 3 , Cl - , and F - were lower for most of the sulfur saturated glasses. The measured Cr 2 O 3 and K 2 O concentrations were markedly lower for the sulfur saturated versions of the glasses as compared to the quenched versions, which may indicate that these elements partitioned to the soluble sulfur salts. A set of modified glasses was received from PNNL later in this study. Chemical analysis showed that these glasses generally met their targeted compositions with the exception of Cr 2 O 3 concentrations, which were considerably higher than the targeted values. The measured SO 3 concentrations were higher for most of the sulfur saturated glasses relative to those of the quenched versions, as expected. Chemical analysis of the wash solutions that resulted from preparation of the sulfur saturated melts identified Cr, K, Na, P, S, and V as the major elements in solution. With the exception of sulfur, this may explain why the measured concentrations of these components in the study glasses were generally below the targeted values. PNNL performed PCTs on quenched and canister centerline cooled (CCC) versions of the study glasses, as well as the modified glasses. The leachates were sent to SRNL for chemical analysis. Minor scatter in the measured values among the triplicate samples for each glass was noted, with significant differences among the triplicate values noted for four of the CCC glasses. Normalized concentration values for some of the study glasses were greater than the 100 g/L maximum (total dissolution), which may indicate an error in the test. It is recommended that PNNL review their performance of the PCTs to determine the cause of these potential outliers and errors.

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

Characterization Results of SRS Tank 26 Samples FTF-26-19-12 and -13

Savannah River Remediation (SRR) has requested that the Savannah River National Laboratory (SRNL) characterize several samples from the Savannah River Site (SRS) Tank Farm in preparation for Sludge Batch (SB) 10. This report documents the characterization of Tank 26 material received by SRNL in May of 2019. SRR pulled 2 x 200 mL samples (FTF-26-19-12 and FTF-26-19-13) with Tank 26 slurry pumps at 60 inches. Two samples were taken to ensure SRNL has adequate material for the requested analyses. Because both samples were taken at the same time and at the same tank level, SRNL combined the two Tank Farm samples into one sample prior to analyses. Results presented in this report include: weight percent solids, density, elemental composition of supernate and slurry, Cs-137 in supernate and slurry, and slurry total alpha and beta. The sample contained higher than expected total sulfur with a significant insoluble fraction, prompting SRR-E to request a washing study to predict how sulfur will be removed during Tank Farm washing.

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

Measurement Acceptance Region (MAR) Assessment Results Based on the October 2019 Sludge Batch 10 Projections

Recent analysis of a Tank 26 sample indicated a higher than anticipated sulfur concentration in the supernatant and a significant amount of insoluble sulfur. The final concentration in the glass product would be well above the sulfate limits of 0.6-0.65 weight percent (wt.%) determined for previous sludge batches. In preparation for Sludge Batch 10 (SB10) processing, washing of Tank 51 will be performed to reduce the sulfate concentration in the glass product to less than 0.65 wt.% at waste loadings (WLs) of interest. To evaluate the feasibility of two Tank 51 Na wash endpoints (0.8 and 1.2M), Savannah River Remediation requested that the Savannah River National Laboratory perform assessments using the Defense Waste Processing Facility Product Composition Control System (PCCS) models and their associated Measurement Acceptance Region (MAR) constraints.

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

Characterization of Hanford LAW Phase 3 Glasses

This report provides analyses of glass compositions, wash solution compositions, and Product Consistency Test (PCT) leachate compositions for a series of simulated low-activity waste (LAW) glasses fabricated at the Pacific Northwest National Laboratory (PNNL). These data will be used in the development of improved property/composition models for LAW glass at Hanford. Chemical analyses were performed on a representative sample of each of the quenched and sulfur saturated glasses to allow for comparisons with the targeted compositions. For some of the quenched glasses, measured concentrations of Cr₂O₃, K₂O, Na₂O, P₂O₅, and ZrO₂ were below the targeted values. For some of the sulfur saturated glasses, the measured concentrations of Al₂O₃, Cl⁻, Cr₂O₃, F⁻, K₂O, Na₂O, P₂O₅, V₂O₅, and ZrO₂ were generally low relative to the targeted values. A comparison of the measured compositions of the quenched and sulfur saturated versions of the study glasses showed that the measured concentrations of Al₂O₃, Cl⁻, and F⁻ were lower for most of the sulfur saturated glasses. The measured Cr₂O₃ and K₂O concentrations were markedly lower for the sulfur saturated versions of the glasses as compared to the quenched versions, which may indicate that these elements partitioned to the soluble sulfur salts. A set of modified glasses was received from PNNL later in this study. Chemical analysis showed that these glasses generally met their targeted compositions. The measured SO₃ concentrations were higher for most of the sulfur saturated glasses relative to those of the quenched versions, as expected. Chemical analysis of the wash solutions that resulted from preparation of the sulfur saturated melts identified Cr, K, Na, P, S, and V as the major elements in solution. With the exception of sulfur, this may explain why the measured concentrations of these components in the study glasses were generally below the targeted values. PNNL performed PCTs on quenched and canister centerline cooled (CCC) versions of the study glasses, as well as the modified glasses. The leachates were sent to SRNL for chemical analysis. Minor scatter in the measured values among the triplicate samples for each glass was noted. Several of the study glasses, both quenched and CCC, have NCi values that are greater than the immobilized LAW constraint of 4 g/L for B, Na, and Si.

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

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Molybdenum Recovery from Filters Used in Large-scale Dissolution of Sintered Mo-disks

99 Mo production with linear accelerators can be achieved via the bremsstrahlung photonuclear reaction 100 Mo(γ,n) 99 Mo or the neutron capture reaction 98 Mo(n,γ) 99 Mo. For commercial producers, maximum recovery of enriched 98 Mo and 100 Mo target material is critical for sustaining an economic production cycle. During the peroxide dissolution of Mo metal disks and subsequent conversion to K 2 MoO 4 , several grams of Mo can be lost during filtration from solution when several hundred grams of sintered Mo disks are processed. This investigation shows that 5–8 g Mo is routinely retained on the filter units, but it can be almost fully recovered using aqueous washes. Washing can be done immediately and incorporated into the dissolution procedure, or it can occur several months after the initial filtration process to decrease processing time.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Nanoporous weakly coordinating anionic frameworks (Final Technical Report)

Under this award, we synthesized and characterized new weakly coordinating, anionic covalent organic frameworks. The motivation was to expand the class of covalent organic frameworks though the introduction of the first weakly coordinating anionic ones. The significance of the weakly coordinating anionic character of these frameworks is associated with the character of the counter (metal) cations in their pores. Those counter-cations, due to the inability to effectively coordinate to the pore walls, have a “naked”, and thus highly electrophilic, Lewis acidic character. Such properties are expected to translate to a number of interesting physical properties that can be potentially used for a range of energy-relevant applications. One of these applications is gas adsorption, separation, and storage. The electrophilic, Lewis acidic character of the counter-cations are expected to strongly bind to gas molecules entering the pores leading to higher gas sorption capacities. Other applications include heterogeneous catalysis, ion exchange, and solid state ion conduction. Within the project we explored a range of chemical blocks, and tested synthesis conditions to polymerize these building blocks to form the target materials. In particular, we focused on materials with group 15 elements like phosphorus in the oxidation state +5 and the coordination number 6, that have a formal -1 charge. The synthesis of such materials is challenging because its success depends on the absence of residual monomeric and oligomeric side-products in the pores, and the absence of significant framework interpenetration. Crystalline forms of these materials require sufficient reversibility of the polymerization reaction (which is difficult to achieve because group 15 elements like P tend to make strong covalent bonds with neighbored elements), as well as suitable polymerization rates that prevent the formation of kinetically controlled products. We had success with the preparation of anionic porous organic frameworks with anionic building blocks containing P(V)O6 octahedra synthesized through Yamamoto coupling reactions. These materials are non-crystalline because of the irreversible nature of the Yamamoto coupling reaction, but show porosity. The porosity and surface area can be significantly increased by co-reaction with electrostatically neutral co-reactants. Very high porosity and surface areas can be achieved when the materials are washed with hydrochloric acid. This method is commonly used to free products from Yamamoto coupling reactions from by-products. Indeed, this method removed by products from the pores of our materials, however it also led to the hydrolysis of the PO6 octahedra to form pending phosphonic acid groups (-PO3H)-groups in the frameworks (PA-POFs) which was actually not desired. Nonetheless, these materials showed high CO2 sorption capacities. In addition, they show extremely high (actually record) sorption capacities for the removal of a range of common micropollutants such as bisphenol A (BPA) and 4-nitrophenol in water. BPA is a wide-spread organic pollutant and a known endocrine disruptor. The maximum adsorption capacity of BPA at equilibrium is found to be as high as 3,366 mg g-1 by Langmuir adsorption model, which is more than 10 times greater than peer materials. The polymer also rapidly removes various other organic micropollutants with more than 90% removal efficiencies. In addition, the PA-POF material can be regenerated at least five times by mild washing using methanol without significant loss in removal efficiency. This is significant advantage over commonly used activated carbon materials (that are in Brita water faucet filters, for example). Comparison studies showed that only materials with the phosphonate groups show extreme sorption although the number of PA group is actually small according to solid state NMR and EDX data. More research would be needed to investigate why this is (a possibility is that the phosphonate groups have a multiplying effect in which a PA-induced initial adsorption leads to the adsorption of much more adsorbate). The extreme sorption capabilities can also be visually seen when colored adsorbates such as 4-nitrophenol are used, as the adsorption of the nitrobenzene leads to the decoloration of the solution. We have produced three short videos that illustrate the sorption capabilities compared to two peer materials (activated carbon from dismantled Brita water filters, and a electrostatically neutral, chemically related, porous organic polymer, respectively).

36 MATERIALS SCIENCE↗
DOE OSTI2021

Characterization of the Sulfur-Saturated Melt Versions of the LAW Phase 4 Glasses

In this report, results are provided from the analyses of a series of sulfur-saturated melt versions of simulated nuclear waste glasses, and chemical analysis of the wash solution resulting from the preparation of these glasses. The glasses were selected and fabricated by the Pacific Northwest National Laboratory as part of a broader study of the influence of glass composition on chemical durability, sulfur retention, and other properties. The resulting data will be used in the development of enhanced property/composition models for waste vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the sulfur-saturated versions of the quenched glasses to allow for comparisons with the targeted compositions, as well as the measured compositions of the quenched glasses. The measured concentrations of chlorine, and fluorine were below the targeted values for most of the study glasses, likely because of volatility during the multiple melting steps. The measured concentrations of B 2 O 3 , K 2 0, and Li 2 0 were generally low relative to the targeted values. As expected, the measured concentrations of SO 3 in most of the glasses were higher than targeted due to the use of the sulfur saturation method in fabricating these glasses. The wash solutions contained mainly sodium and sulfur, with moderate concentrations of potassium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

SB10 Frit Recommendation, and Evaluations of the Glass Variability Study and Cs-137 Concentrations in Strip Effluent Based on May 2021 Projections

The Defense Waste Processing Facility (DWPF) is currently preparing to initiate processing of Sludge Batch 10 (SB10), which is comprised of material from Tanks 11H, 13H, 15H, and 26F, Alternate Feed Stock-2 (AFS-2) and Sodium Reactor Experiment (SRE) material from H-Canyon. In support of SB10 qualification, frit development using 2020 Tank 40 blend projections and experimental work for the glass variability study were previously conducted. Frit 473 and Frit 209 were identified as candidate frits and both were included in the development of the variability study test matrix; however, a final frit recommendation was postponed until more information could be determined about the composition of SB10 after washing. In May 2021, Savannah River Remediation (SRR) reprojected SB10 based on the analytical results from the Tank 51 qualification sample that was washed in the Savannah River National Laboratory (SRNL) Shielded Cells Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Selection of Glasses to Confirm the 0.65 Weight Percent Sulfate Solubility Limit for Sludge Batch 10

In preparation for Sludge Batch 10 (SB10) processing, projections of sulfate (SO 4 2- ) in glass at 36% waste loading (WL) were calculated in May 2020 for Tank 40 blend projections representing 0.7M and 0.85M Na wash endpoints. The projected SO 4 2- ) concentrations for either sludge-only (SO) or coupled processing with the Salt Waste Processing Facility (SWPF) were either near or exceeded the current Sludge Batch 9 (SB9) limit of 0.65 weight percent (wt.%). Four nominal glass compositions were selected based on SO and coupled processing for the 0.85M Na wash endpoint Tank 40 blend projection to conduct an initial evaluation of the SB10 sulfate solubility behavior. A sulfate salt phase was absent from each of the prepared glasses, which provided preliminary results that supported the 0.65 wt.% SO 4 2- ) limit for SB10.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

X-ray diffraction and scanning electron microscope characterizations of the salt waste processing facility tank 101 and tank 104 solid samples

The “as-received” Salt Waste Processing Facility (SWPF) Tank 101 and two Tank 104 samples (pre-wash and washed solids) have been characterized using x-ray diffraction (XRD) and scanning electron microscope/energy dispersive x-ray (SEM/EDX) techniques to address the concerns for the existence of solids, which may interfere with the operational performance of the SWPF crossflow filtration process.

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