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At least 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Analytical Results for MOX Colemanite Concrete Samples Received on January 15, 2013

The Mixed Oxide Fuel Fabrication Facility (MFFF) will use colemanite bearing concrete neutron absorber panels credited with attenuating neutron flux in the criticality design analyses and shielding operators from radiation. The Savannah River National Laboratory (SRNL) is tasked with measuring the total density, partial hydrogen density, and partial boron density of the colemanite concrete. SRNL received twelve samples of colemanite concrete for analysis on January 15, 2013. The average total density of each of the samples measured by the ASTM method C 642, the average partial hydrogen density was measured using method ASTM E 1131, and the average partial boron density of each sample was measured according to ASTM C 1301. The lower limits and measured values for the total density, hydrogen partial density, and boron partial density are presented. For all the samples tested, the total density and the hydrogen partial density met or exceeded the specified limit. All of the samples met or exceeded the boron partial density lower bound with the exception of samples G3-M11-2000-H, G3-M11-3000-M, and G5-M1-3000-H which are below the limit of 1.65E-01 g/cm 3 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Assessment of the LV-S2 & LV-S3 Stack Sampling Probe Locations for Compliance with ANSI/HPS N13.1-1999

This document reports on a series of tests conducted to assess the proposed air sampling locations for the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Group 1-2A exhaust stacks with respect to the applicable criteria regarding the placement of an air sampling probe. The LV-C2, LV-S2, and LV-S3 exhaust stacks were tested together as a group (Test Group 1-2A). This report only covers the results of LV-S2 and LV-S3; LV-C2 will be reported on separately. Federal regulations1 require that a sampling probe be located in the exhaust stack according to the criteria established by the American National Standards Institute/Health Physics Society (ANSI/HPS) N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stack and Ducts of Nuclear Facilities. 2 These criteria address the capability of the sampling probe to extract a sample that represents the effluent stream.

Glissmeyer, John A.↗
DOE OSTI2015

Results of Hg Speciation Testing on 2Q15 Tank 50 and Tank 21 TCLP Extraction Fluid Samples and Tank 49 Material

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The sixth shipment of samples was designated to include SRNL Toxicity Characteristic Leaching Procedure (TCLP) extraction fluids from the 2Q15 Tank 50 and Tank 21 saltstone waste forms as well as Tank 49 material (HTF-49-15-45) collected on May 7, 2015. The TCLP extraction fluid samples were collected from duplicate extractions of two crushed saltstone fractions, one designated as “large” and one designated as “normal”. Both fractions of 2Q15 Tank 50 and Tank 21 saltstones (eight samples) were collected with the minimum possible free headspace and refrigerated prior to final dilution. The Tank 49 sample was left in its stainless steel dip bottle until it was opened and an aliquot diluted 1:100 with Eurofins deionized water on May 21, 2015. Following initial dilution in the Shielded Cells, the diluted Tank 49 sample was immediately moved to refrigeration and kept in the dark until prepared for shipment to Eurofins.

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

Analysis Of Condensate Samples In Support Of The Antifoam Degradation Study

The degradation of Antifoam 747 to form flammable decomposition products has resulted in declaration of a Potential Inadequacy in the Safety Analysis (PISA) for the Defense Waste Processing Facility (DWPF). Savannah River National Laboratory (SRNL) testing with simulants showed that hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and 1-propanal are formed in the offgas from the decomposition of the antifoam. A total of ten DWPF condensate samples from Batch 735 and 736 were analyzed by SRNL for three degradation products and additional analytes. All of the samples were analyzed to determine the concentrations of HMDSO, TMS, and propanal. The results of the organic analysis found concentrations for propanal and HMDSO near or below the detection limits for the analysis. The TMS concentrations ranged from below detection to 11 mg/L. The samples from Batch 736 were also analyzed for formate and oxalate anions, total organic carbon, and aluminum, iron, manganese, and silicon. Most of the samples contained low levels of formate and therefore low levels of organic carbon. These two values for each sample show reasonable agreement in most cases. Low levels of all the metals (Al, Fe, Mn, and Si) were present in most of the samples.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

ANALYSIS OF CONDENSATE SAMPLES IN SUPPORT OF THE ANTIFOAM DEGRADATION STUDY

The degradation of Antifoam 747 to form flammable decomposition products has resulted in declaration of a Potential Inadequacy in the Safety Analysis (PISA) for the Defense Waste Processing Facility (DWPF). Savannah River National Laboratory (SRNL) testing with simulants showed that hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and 1-propanal are formed in the offgas from the decomposition of the antifoam. A total of ten DWPF condensate samples from Batch 735 and 736 were analyzed by SRNL for three degradation products and additional analytes. All of the samples were analyzed to determine the concentrations of HMDSO, TMS, and propanal. The results of the organic analysis found concentrations for propanal and HMDSO near or below the detection limits for the analysis. The TMS concentrations ranged from below detection to 11 mg/L. The samples from Batch 736 were also analyzed for formate and oxalate anions, total organic carbon, and aluminum, iron, manganese, and silicon. Most of the samples contained low levels of formate and therefore low levels of organic carbon. These two values for each sample show reasonable agreement in most cases. Low levels of all the metals (Al, Fe, Mn, and Si) were present in most of the samples.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Radiological results for samples collected on paired glass- and cellulose-fiber filters at the Sandia complex, Tonopah Test Range, Nevada

Airborne particulates are collected at U.S. Department of Energy sites that exhibit radiological contamination on the soil surface to help assess the potential for wind to transport radionuclides from the contamination sites. Collecting these samples was originally accomplished by drawing air through a cellulose-fiber filter. These filters were replaced with glass-fiber filters in March 2011. Airborne particulates were collected side by side on the two filter materials between May 2013 and May 2014. Comparisons of the sample mass and the radioactivity determinations for the side-by-side samples were undertaken to determine if the change in the filter medium produced significant results. The differences in the results obtained using the two filter types were assessed visually by evaluating the time series and correlation plots and statistically by conducting a nonparametric matched-pair sign test. Generally, the glass-fiber filters collect larger samples of particulates and produce higher radioactivity values for the gross alpha, gross beta, and gamma spectroscopy analyses. However, the correlation between the radioanalytical results for the glass-fiber filters and the cellulose-fiber filters was not strong enough to generate a linear regression function to estimate the glass-fiber filter sample results from the cellulose-fiber filter sample results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

2015 Long-Term Hydrologic Monitoring Program Sampling and Analysis Results at Rio Blanco, Colorado

The U.S. Department of Energy (DOE) Office of Legacy Management conducted annual sampling at the Rio Blanco, Colorado, Site for the Long-Term Hydrologic Monitoring Program (LTHMP) on May 20–21, 2015. This report documents the analytical results of the Rio Blanco annual monitoring event, the trip report, and the data validation package. The groundwater and surface water monitoring samples were shipped to the GEL Group Inc. laboratories for conventional analysis of tritium and analysis of gamma-emitting radionuclides by high-resolution gamma spectrometry. A subset of water samples collected from wells near the Rio Blanco site was also sent to GEL Group Inc. for enriched tritium analysis. All requested analyses were successfully completed. Samples were collected from a total of four onsite wells, including two that are privately owned. Samples were also collected from two additional private wells at nearby locations and from nine surface water locations. Samples were analyzed for gamma-emitting radionuclides by high-resolution gamma spectrometry, and they were analyzed for tritium using the conventional method with a detection limit on the order of 400 picocuries per liter (pCi/L). Four locations (one well and three surface locations) were analyzed using the enriched tritium method, which has a detection limit on the order of 3 pCi/L. The enriched locations included the well at the Brennan Windmill and surface locations at CER-1, CER-4, and Fawn Creek 500 feet upstream.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Analytical Results from Routine DSSHT and SEHT Monthly Samples

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the “microbatches” of Integrated Salt Disposition Project (ISDP) Salt Batch (“Macrobatch”) 8B have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (Inductively Coupled Plasma Emission Spectroscopy - ICPES), and anions (Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatch samples. The Actinide Removal Process (ARP) and the Modular Caustic-Side Solvent Extraction Unit (MCU) continue to show more than adequate Pu and Sr removal for times when monosodium titanate (MST) is used. Even with no MST strike being performed there exists some small Pu and Sr removal, likely from filtration of fines containing these elements. The Cs removal continues to be excellent, with decontamination factors (DF) averaging 16,400. The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior. SRNL recommends that a sample of the strip feed be analyzed for cation and anion content if a further decline in boron concentration is noted in future SEHT samples.

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DOE OSTI2016

Analysis of Selected Legacy 85 Kr Samples

Legacy samples composed of 85 Kr encapsulated in solid zeolite 5A material and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing have been preserved. The samples were a result of krypton R&D encapsulation efforts in the late 1970s performed at the Idaho Chemical Processing Plant. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. Upon receipt the outer shipping package was opened, and the inner package, removed and placed in a radiological hood. The individual capsules were double bagged as they were removed from the inner shipping pig and placed into individual glass sample bottles for further analysis. The five capsules were then x-ray imaged. Capsules 1 and 4 appear intact and to contain an amorphous mass within the capsules. Capsule 2 clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. Capsule 3 shows similar bead-like material within the intact capsule. Capsule 5 had been opened at an undetermined time in the past. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of Capsule 5. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the nondestructive analysis (NDA) results, a proposed path forward was formulated to advance this effort toward the original goals of understanding the effects of extended storage on the waste form and package. Based on the initial NDA and the fact that there are at least two breached samples, it was proposed that exploratory tests be conducted with the breached specimens before opening the three intact capsules. Portions of these would be analyzed to determine the fraction of krypton/xenon remaining in the matrix and the amount of rubidium remaining in the matrix. The inner surface of the breached capsules would be examined for corrosion. The materials contained in Capsules 2 and 5 have been examined. There appears to be a relatively uniform distribution of Kr and Rb throughout the pellets examined. The chemical composition of the pellets appears to be consistent with 5A molecular sieves. The material contained within Capsule 5 showed ~1 at. % lead. The origin of the Pb is currently indeterminate. X-ray diffraction analysis shows a significant shift from the 5A structure, most likely due to the Kr encapsulation / sintering process that occurred when the samples were made. The capsule walls were also examined and showed extensive corrosion throughout. Elemental mapping of the capsule material appeared consistent with carbon steel, while the weld material appeared consistent with a stainless steel. The interior surface of the capsule appeared to have a layer of material containing Al, Si, and Ca similar to the 5A molecular sieve. Analysis for Rb within the corrosion sites was inconclusive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Data Validation Package, December 2015, Groundwater Sampling at the Bluewater, New Mexico, Disposal Site, September 2016

Groundwater samples were collected from monitoring wells at the Bluewater, New Mexico, Disposal Site to monitor groundwater contaminants as specified in the 1997 Long-Term Surveillance Plan for the DOE Bluewater (UMTRCA Title II) Disposal Site Near Grants, New Mexico (LTSP). Sampling and analyses were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PRO/S04351, continually updated). A duplicate sample was collected from location HMC-951. Alluvium wells are completed in the alluvial sediments in the former channel of the Rio San Jose, which was covered by basalt lava flows known as the El Malpais, and are identified by the suffix (M). Bedrock wells are completed in the San Andres Limestone/Glorieta Sandstone hydrologic unit (San Andres aquifer) and are identified by the suffix (SG). Wells HMC-951 and OBS-3 are also completed in the San Andres aquifer. The LTSP requires monitoring for molybdenum, selenium, uranium, and polychlorinated biphenyls (PCBs); PCB monitoring occurs only during November sampling events. This event included sampling for an expanded list of analytes to characterize the site aquifers and to support a regional groundwater investigation being conducted by the New Mexico Environment Department.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Sampling for Air Chemical Emissions from the Life Sciences Laboratory II

Sampling for air chemical emissions from the Life Science Laboratory II (LSL-II) ventilation stack was performed in an effort to determine potential exposure of maintenance staff to laboratory exhaust on the building roof. The concern about worker exposure was raised in December 2015 and several activities were performed to assist in estimating exposure concentrations. Data quality objectives were developed to determine the need for and scope and parameters of a sampling campaign to measure chemical emissions from research and development activities to the outside air. The activities provided data on temporal variation of air chemical concentrations and a basis for evaluating calculated emissions. Sampling for air chemical emissions was performed in the LSL-II ventilation stack over the 6-week period from July 26 to September 1, 2016. A total of 12 sampling events were carried out using 16 sample media. Resulting analysis provided concentration data on 49 analytes. All results were below occupational exposure limits and most results were below detection limits. When compared to calculated emissions, only 5 of the 49 chemicals had measured concentrations greater than predicted. This sampling effort will inform other study components to develop a more complete picture of a worker’s potential exposure from LSL-II rooftop activities. Mixing studies were conducted to inform spatial variation in concentrations at other rooftop locations and can be used in conjunction with these results to provide temporal variations in concentrations for estimating the potential exposure to workers working in and around the LSL-II stack.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Characterization of Tank 51 Sludge Slurry Samples (HTF-51-17-44/ HTF-51-17-48) in Support of Sludge Batch 10 Processing

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) Engineering (SRR-E) to provide sample characterization and analyses of Tank 51 sludge samples in support of Sludge Batch (SB) 10. The two Tank 51 sludge samples were sampled and delivered to SRNL in May of 2017. These two Tank 51 sludge samples were combined into one composite sample and analyzed for corrosion controls analytes, select radionuclides, chemical elements, density and weight percent total solids and aluminum hydroxides (gibbsite and boehmite) by x-ray diffraction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Analysis of Sludge Batch 10 Samples from Tank 51H (HTF-51-18-94, -95)

Savannah River National Laboratory characterized sludge samples from Tank 51H that is being used to generate Sludge Batch 10. The composite slurry sample from Tank 51H was analyzed and determined to contain a concentrated salt solution with a sodium concentration of 5.79 M (1.33E+05 mg/L) and insoluble sludge solids content of 7.20 wt% (using all nine replicates from the 3 wt% total solids measurements). Hydroxide is the main anionic species in the salt solution with a concentration of 3.00 M followed by nitrate (0.728 M) and nitrite (0.657 M). The sum of the major cations versus the sum of the major anions from the analysis of the filtered supernatant liquid phase shows a difference of <5% indicating good data quality. The results from the analysis of the digestions of the well mixed composite slurry sample show sodium to be the main component of the total dried solids of the sample followed by aluminum, iron, manganese, and thorium. Additional measurements of the weight percent total solids and density using hand shaking of the sample bottle and recirculation through a peristaltic pump indicate both methods provide adequate mixing of the sludge slurry sample.

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.

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DOE OSTI2021

Analysis of Tank 38H (HTF-38-20-103, -104) and Tank 43H (HTF-43-20-105, -106) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. The results indicate the concentrations of most species in the Tank 38H surface sample decreased from the previous surface sample. The Tank 38H sub-surface sample shows only minor changes in concentration for soluble species in the solution (e.g., Na, Al, Cs-137) but a large drop in concentrations for species typically associated with sludge solids (e.g., U, Pu, Fe, Si) likely because of a decrease in sludge solids from the previous sample. The large differences in the concentrations of major components between the Tank 38H surface and sub-surface samples indicate significant stratification of solution species within the tank.

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

Underground Test Area (UGTA) Sampling Plan for Corrective Action Units 101 and 102: Central and Western Pahute Mesa, Nevada National Security Site, Nevada

The purpose of this groundwater sampling plan (referred to herein as “the Plan”) is to describe the approach for collecting and analyzing groundwater samples to meet the objectives of the U.S. Department of Energy (DOE), Environmental Management (EM) Nevada Program’s Underground Test Area (UGTA) Activity. The Plan is focused on the Central and Western Pahute Mesa corrective action units (CAUs 101 and 102). The remaining UGTA CAUs are in the final stage of the UGTA strategy (Closure Report [CR] stage), and monitoring requirements for these CAUs are described in their respective CRs. The Plan supersedes the Nevada National Security Site Integrated Groundwater Sampling Plan and its ensuing Record of Technical Change (ROTC) (DOE/EMNV, 2019b). The Plan is designed to ensure compliance with the UGTA Quality Assurance Plan (QAP) and its ensuing ROTC (DOE/EMNV, 2018), and the Federal Facility Agreement and Consent Order (FFACO) (1996, as amended). The Plan is implemented by the Environmental Program Services (EPS) contractor. Implementation of the Plan provides data to complete corrective actions prescribed by the FFACO to protect the public from groundwater contaminated by historical underground nuclear testing. The Plan’s scope comprises sample collection and analysis requirements for assessing the extent of groundwater contamination from underground nuclear testing. Sampling wells located in the communities surrounding the Nevada National Security Site (NNSS) is performed by other organizations (e.g., management and operating [M&O] contractor) and is therefore not included within the Plan. The results of the community sampling are presented in the annual NNSS environmental reports.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Optimized Microwave Digestion and Quantification Procedure for Boron Carbide samples

A microwave-assisted HNO3-H2SO4-HF digestion system was explored/optimized for the total dissolution of Boron Carbide samples followed by multi-elemental determination using ICP-MS and ICP OES, in order to improve the methods used in a previous try using a microwave-assisted H2O2-HNO3-HF digestion system. The samples were microwave digested to accomplish complete dissolution needed to perform quantitative analysis of their metal content. Two Boron Carbide samples were provided to us: JM10 and JM11. The Boron Carbide samples were completely dissolved at the end of the optimized microwave digestion procedure that was developed. The digested samples were analyzed for metals mostly by ICP MS; few elements were analyzed by ICP-OES. Microwave digestion reactions were performed in a Titan MPS instrument (Perkin Elmer, USA). Titan MPS instrument is equipped with Temperature and Pressure regulations and controlled by software. Each microwave digestion batch contained the samples of interest along with a reagent blank and one certified standard (1632e) in order to (i) compensate for the contaminations present in the reagents, (ii) ensure that the digestion was complete and (iii) that there was a good recovery of all the constituents. The certified standard used (1632e) was provided with USGS certificates of analysis.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Cesium Batch Contact Equilibrium Testing of Crystalline Silicotitanate (CST) Sieve Cuts in SRS Average Simulant and Examination of CST Samples Aged in Caustic and Simulant in Support of Tank Closure Cesium Removal 1A (TCCR-1A)

Batch contact testing to determine cesium equilibrium loading on Crystalline Silicotitanate ion exchange media in Savannah River Site Average Waste Simulant at 25 °C indicated that smaller diameter particles isolated by sieving pretreated CST media may load slightly higher (<10%) amounts of cesium, though the differences are within analytical uncertainty. In addition, ion exchange media sub-samples stored in 2-4 M NaOH and caustic simulant solutions for ~2.5 years were examined by optical microscopy and the 3 M NaOH sample was also analyzed to determine whether changes in the particle size distribution occurred during storage. No visual indications of particle attrition or agglomeration were observed for any sample. Particle size analysis indicated that a slight decrease occurred in the average particle diameter following contact with 3 M NaOH (541 µm average diameter versus 566 µm for the pretreated CST prior to contact). A small increase (from 0 to <0.5 wt. %) in the number of particles ranging from 271 and 322 µm was also observed for the CST sample contacted with 3 M NaOH relative to a sample of the original pretreated material. However, this small change could be due to sub-sampling differences or analytical uncertainty. It does not appear that small particles are formed to a significant degree during CST caustic contact or that small particles which do form (presumably from attrition of larger particles during pretreatment) load significantly more cesium than the bulk material. Minimal other negative consequences were observed associated with CST extended caustic or simulant contact, except for the tendency for more concentrated salt solutions to form some salt crystals which deposit on the media over time.

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