Defense Waste Processing Facility Research Articles

Analysis of antifoam agent degradation products in an evaporator

ABSTRACTA wetting agent used to control foaming in the Chemical Processing Cell at the Defense Waste Processing Facility degrades to form compounds that could volatilize to form vapor exceeding the lower flammability limit. Three identified components of concern were hexamethyl disiloxane, trimethyl silanol, and propanal. Analytical methods were developed and implemented on a real waste sample to monitor degradation products. Using standards, an extraction method with dichloromethane and analysis by gas chromatography-mass spectrometry and hydrogen nuclear magnetic resonance was developed. Both methods had detection limits less than 1 mg/L for the analytes.

Determination of the Impact of Glycolate on Cs, Sr, and Actinide Separations in High-Level Waste

The Defense Waste Processing Facility (DWPF), responsible for vitrifying high-level waste (HLW) at the Savannah River Site (SRS), is planning to introduce glycolic acid as a reductant during chemical conditioning of the waste prior to vitrification. A portion of this glycolic acid may be recycled back to the tank farm as sodium glycolate. Since glycolate can serve as a complexing agent for a variety of metallic ions, recent studies were performed to examine the effect of glycolate on the removal of Sr and actinides by monosodium titanate (MST) and modified MST (mMST) and the separation of Cs by the caustic-side solvent extraction (CSSX) process.

High Chrome Refractory Characterization: Part II. Accumulation of Spinel Corrosion Deposits in Radioactive Waste Glass Melters

High Cr2O3 containing Monofrax™ K‐3 is a robust refractory that is used in radioactive waste glass melters worldwide. Monofrax™ K‐3 contains highly reduced phases. Conversely, many of the radioactive feeds being processed are highly oxidizing. The K‐3 refractory corrosion rates in oxidizing (high nitrate) feeds were ~1.8–2.8 times higher than the rates determined using reducing feeds. The corrosion product formed is a mixture of spinel and glass (slag) that can accumulate on the melter floor. A methodology to calculate the depth of slag deposits from refractory corrosion is presented and verified with slag measurements from the Defense Waste Processing Facility (DWPF) melter after it had processed oxidized feeds for 1.75 years. The calculations show that had the facility continued to process oxidized feeds the melter lifetime (based on when the deposits could have reached and blocked the pour spout riser) would have been ~4.5 years. The DWPF changed to a reducing flow sheet after ~3 years of operation. The lifetimes of Melter #1 and Melter #2, assuming a failure due to pour spout blockage, are calculated as 7.7–12 years based on corrosion rates measured with reducing feeds. Lifetimes of 9 and >11 years have actually been achieved.

DWPF Startup Frit Viscosity Measurement Round Robin Results

A viscosity standard is needed to replace the National Institute of Standards and Technology (NIST) glasses currently being used to calibrate viscosity measurement equipment. The current NIST glasses are either unavailable or less than ideal for calibrating equipment to measure the viscosity of high‐level waste glasses. This report documents the results of a viscosity round robin study conducted on the Defense Waste Processing Facility (DWPF) startup frit. DWPF startup frit was selected because its viscosity‐temperature relationship is similar to most DWPF and Hanford high‐level waste glass compositions. The glass underwent grinding and blending to homogenize the large batch. Portions of the batch were supplied to eight laboratories for viscosity measurements, which were conducted following a specified temperature schedule with a temperature range of 1150°C–950°C, with an option to measure viscosity at temperatures below 950°C if their equipment was capable. Results were used to fit the Vogel‐Fulcher‐Tamman‐Hesse and Arrhenius equations to viscosity as a function of temperature for the entire temperature range of 460°C through 1250°C as well as the limited temperature interval of ~ 950°C through 1250°C. The standard errors for confidence and prediction were determined for the fitted models.

Analysis of the Leaching Efficiency of Inhibited Water and Tank Simulant in Removing Residues on Thermowell Pipes

A cleaning protocol was determined and modeled, using the shrinking core model, for the removal of a solid residue coating thermo well pipes contained in the riser of a Savannah River Site (SRS) waste tank (Tank 48H). The solid residues on two sets of thermo well pipe samples removed from the D2 riser in SRS Tank 48H were characterized by high performance liquid chromatography (HPLC), ion chromatography (IC) and gamma spectroscopy. The residue thickness was determined using the ASTM standard D 3483-05 and was found to be three orders of magnitudes below the 1 mm thickness estimated from an earlier video of the tank cooling coil inspection. The actual estimated thickness ranged from 4 to 20.4 microns. The mass per unit area ranged from 0.15 to 0.82 milligrams per square centimeter. The residues appear to consist primarily of potassium tetraphenylborate (39.8 wt% KTPB) and dried salt solution (33.5 wt% total of nitrates, nitrites, and oxalate salts), although ∼30% of the solid mass was not accounted for in the mass balance. No evidence of residue buildup was found inside the pipe, as expected. Two aqueous solutions representing waste processing streams were chosen for the cleaning trials. The residue leaching characteristics were measured by placing one pipe in inhibited water (0.01 M NaNO2 and 0.01 M NaOH) and one pipe in Defense Waste Processing Facility (DWPF) Recycle simulant (primarily 0.20 M NaNO2 and 0.18 M NaOH). After soaking for less than 4 weeks, the inhibited water was 95.4% effective at removing the residue and the DWPF Recycle simulant was 93.5% effective. The surface appearance of the pipes after leaching tests appeared close to the clean shiny appearance of a new pipe. Total gamma counts of leachates averaged 48.1 dpm/mL, or an equivalent of 2.35E-11 Ci/gm Cs-137 (dry solids basis), which is much lower than the 1.4 E-03 Ci/gm expected for Tank 48H dry slurry solids.

Processing Macrobatch 2 at the Savannah River Site Integrated Salt Disposition Process (ISDP)

The Savannah River Site (SRS) is currently removing liquid radioactive waste from the tanks in its Tank Farm. To treat waste streams that are high in 137Cs, 90Sr, and/or actinides, SRS developed the Actinide Removal Process (ARP) and the Modular Caustic Side Solvent Extraction (CSSX) Unit. Collectively, these two processes make up the Integrated Salt Disposition Process (ISDP). The ARP part is responsible for the removal of strontium and actinides, while the MCU part is responsible for removing cesium. This paper discusses the qualification testing of the second batch of caustic waste that is being processed through ISDP currently. This paper also describes the tests conducted and compares results with current facility performance. The ARP contacts the salt solution with monosodium titanate (MST) to sorb strontium and select actinides. After MST contact, the resulting slurry is filtered to remove the MST (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to ∼5 wt%, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility (DWPF) for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), then strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to DWPF for vitrification. In order to predict waste behavior, the Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples of the second waste batch – Macrobatch 2 – for processing prior to the start of the operation. Testing included MST sorption to remove strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass removal and concentration. This paper describes the tests conducted and compares results from MCU facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Our work indicates that the bench scale tests are a conservative predictor of actual waste performance.

Impact of Eliminating Mercury Removal Pretreatment on the Performance of a High-Level Radioactive Waste Melter Offgas System

Abstract The Defense Waste Processing Facility (DWPF) at the Savannah River Site processes high-level radioactive waste from the processing of nuclear materials that contains dissolved and precipitated metals and radionuclides. Vitrification of this waste into borosilicate glass for ultimate disposal at a geologic repository involves chemically modifying the waste to make it compatible with the glass melter system. Pretreatment steps include removal of excess aluminum by dissolution and washing, and processing with formic and nitric acids to (1) adjust the reduction–oxidation (redox) potential in the glass melter to reduce radionuclide volatility and improve melt rate; (2) adjust feed rheology; and (3) reduce, by steam stripping, the amount of mercury that must be processed in the melter. Elimination of formic acid pretreatment has been proposed to eliminate the production of hydrogen in the pretreatment systems; alternative reductants would be used to control redox. However, elimination of formic acid wo...

Cold Crucible Vitrification of U-bearing SRS SB4 HLW Surrogate

AbstractSavannah River Site Defense Waste Processing Facility (DWPF) Sludge Batch 4 (SB4) high level waste (HLW) simulant at 55 wt % waste loading was produced in the demountable cold crucible and cooled to room temperature in the cold crucible. Appreciable losses of Cs, S and Cl took place during the melting. A second glass sample was subjected to canister centerline cooling (CCC) regime in an alumina crucible in a resistive furnace. X-ray diffraction (XRD) study showed that the glass blocks were composed of vitreous and spinel structure phases. No separate U-bearing phases were found.

Vitrification of High-Level Waste at the Savannah River Site

The objective of this study was to experimentally measure the properties and performance of a series of glasses with compositions that could represent high level waste Sludge Batch 5 (SB5) as vitrified at the Savannah River Site Defense Waste Processing Facility. These data were used to guide frit optimization efforts as the SB5 composition was finalized. Glass compositions for this study were developed by combining a series of SB5 composition projections with a group of candidate frits. The study glasses were fabricated using depleted uranium and their chemical compositions, crystalline contents and chemical durabilities were characterized. Trevorite was the only crystalline phase that was identified in a few of the study glasses after slow cooling, and is not of concern as spinels have been shown to have little impact on the durability of high level waste glasses. Chemical durability was quantified using the Product Consistency Test (PCT). All of the glasses had very acceptable durability performance. The results of this study indicate that a frit composition can be identified that will provide a processable and durable glass when combined with SB5.

Glass Formulation Development in Support of Melter Testing to Demonstrate Enhanced High Level Waste Throughput

AbstractThe U.S. Department of Energy (DOE) is currently processing high-level waste (HLW) through a Joule-heated melter (JHM) at the Savannah River Site (SRS) and plans to vitrify HLW and Low activity waste (LAW) at the Hanford Site. Over the past few years at the Defense Waste Processing Facility (DWPF), work has concentrated on increasing waste throughput. These efforts are continuing with an emphasis on high alumina concentration feeds. High alumina feeds have presented specific challenges for the JHM technology regarding the ability to increase waste loading yet still maintain product quality and adequate throughput. Alternatively, vitrification technology innovations are also being investigated as a means to increase waste throughput. The Cold Crucible Induction Melter (CCIM) technology affords the opportunity for higher vitrification process temperatures as compared to the current reference JHM technology. Higher process temperatures may allow for higher waste loading and higher melt rate.Glass formulation testing to support melter demonstration testing was recently completed. This testing was specifically aimed at high alumina concentration wastes. Glass composition/property models developed for DWPF were utilized as a guide for formulation development. Both CCIM and JHM testing will be conducted so glass formulation testing was targeted at both technologies with a goal to significantly increase waste loading and maintain melt rate without compromising product quality.

Analysis of cesium extracting solvent using GCMS and HPLC

A high-level waste (HLW) remediation process scheduled to begin in 2007 at the Savannah River Site is the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The MCU will use a hydrocarbon solvent (diluent) containing a cesium extractant, a calix[4]arene compound, to extract radioactive cesium from caustic HLW. The resulting decontaminated HLW waste or raffinate will be processed into grout at the Saltstone Production Facility (SPF). The cesium containing CSSX stream will undergo washing with dilute nitric acid followed by stripping of the cesium nitrate into a very dilute nitric acid or the strip effluent stream and the CSSX solvent will be recycled. The Defense Waste Processing Facility (DWPF) will receive the strip effluent stream and immobilize the cesium into borosilicate glass. Excess CSSX solvent carryover from the MCU creates a potential flammability problem during DWPF processing. Bench-scale DWPF process testing was performed with simulated waste to determine the fate of the CSSX solvent components. A simple high performance liquid chromatography (HPLC) method was developed to identify the modifier (which is used to increase Cs extraction and extractant solubility) and extractant within the DWPF process. The diluent and triocytlamine (which is used to suppress impurity effect and ion-pair disassociation) were determined using gas chromatography mass spectroscopy (GCMS). To close the organic balance, two types of sample preparation methods were needed. One involved extracting aqueous samples with methylene chloride or hexane, and the second was capturing the off gas of the DWPF process using carbon tubes and rinsing the tubes with carbon disulfide for analysis. This paper addresses the development of the analytical methods and the bench-scale simulated waste study results.

Cold Crucible Vitrification of Defense Waste Surrogate and Vitrified Product Characterization

ABSTRACTIn the framework of the contract “Advanced Melter Technology Application to the Defense Waste Processing Facility (DWPF) –Cold Crucible Induction Heated Melter (CCIM)”, vitrification tests with Savannah River Site defense waste surrogate were performed at the SIA Radon facility. Cold crucible melters with inner diameter of 216 mm and 418 mm were used in the testing. Commercially available (USA) frits 200 and 320 were used as glass-forming additives. In three different test campaigns, waste additive mixtures were fed as slurries with ∼60 wt.%, ∼30 wt.%, and 45 wt.% water content. Maximum slurry capacity and glass productivity under steady-state conditions were 35.4 kg/h and 16.2 kg/h, respectively. Specific glass productivity reached up to ∼3000 kg/(m2×day). The average melt process temperature was 1250- 1350 °C. Waste loadings in glass were 45 wt.% in tests 1 and 2 and 50 wt.% in test 3. The glasses produced were found to be homogeneous but contained a magnetite-type phase with the spinel structure due to high iron and manganese content in waste. Spinel was observed in the glassy matrix as individual regular crystals and their aggregates. All the waste uranium entered the vitreous phase. Infra-red spectra consist of strong absorption bands due to bridging Si-O-Si and non-bridging Si-O- bonds, some weak bands due to B-O bonds, and a number of narrow bands due to occurrence of the crystalline phase. The glassy products demonstrate high leach resistance. Normalized release of major glass elements (Na, Li, B, Si) is by 10 to 50 times lower than the values required for repository disposition by EPA.

Neptunium Disposal to the Savannah River Site Tank Farm

:Researchers investigated the neutralization of an acidic neptunium solution from a Savannah River Site (SRS) processing canyon and the properties of the resulting slurry to determine the feasibility of disposal in the SRS tank farm. The acidic solution displayed no properties that precluded the proposed disposal route. Neutralization of the acidic neptunium forms a 4wt% slurry of precipitated metal hydroxides. The insoluble solids consist largely of iron (92%) and neptunium hydroxides (2%). The concentration of soluble neptunium remaining after neutralization equaled 1.8 + 0.6mg/L, much less than previous solubility measurements predicted. Researchers used an apparatus similar to an Ostwald-type viscometer to estimate the consistency of the neptunium slurry with the solids present. The yield stress and consistency of the 4 wt% slurry will allow transfer through the tank farm, although concentration of the insoluble solids above 4wt% may cause significant problems due to increased consistency and yield stress. The consistency of the 4wt% slurry is 7.6 centipoise (cP) with a yield stress less than 1 Pascal (Pa). The neptunium slurry, when combined with actual washed radioactive sludge, slightly reduces the yield stress and consistency of the sludge and produces a combined slurry with acceptable rheological properties for vitrification in the Defense Waste Processing Facility (DWPF).

Removal of Silicon from High-Level Waste Streams via Ferric Flocculation

:The presence of silicate from glass-forming frit in the recycle waste from the Defense Waste Processing Facility produces wastes that, when combined with the traditional aluminate-bearing wastes stored in Savannah River Site’s tank farms, can produce insoluble sodium aluminosilicates. Currently, aluminum-bearing wastes and silicon-bearing wastes are processed in separate evaporators. This, however, limits operational flexibility. Therefore, treatment to remove silicon has been proposed to allow greater flexibility for processing these wastes in the site’s evaporators. The use of a ferric precipitation (flocculation) to remove the silicon has been tested using waste simulants.Ferric precipitation following ferric nitrate addition to two different alkaline DWPF recycle waste simulants was effective at removing silicon to levels below concern for processing in any of the Savannah River Site’s evaporators. Removal of silicon was rapid with removal complete in 2h. Elevated temperatures were tested and found to be not required. Capacities of the resultant iron flocculent were ~0.1 g of silicon per gram of iron at a final silicon concentration of 50mg/L. Moreover, the silicon removal obeys a classical Freundlich adsorption isotherm.

Field Comparison of the Sampling Efficacy of Two Smear Media: Cotton Fiber and Kraft Paper

Two materials were compared in field tests at the Defense Waste Processing Facility: kraft paper (a strong, brown paper made from wood pulp prepared with a sodium sulfate solution) and cotton fiber. Based on a sampling of 46 pairs of smears, the cotton fiber smears provide a greater sensitivity. The cotton fiber smears collected an average of 44% more beta activity than the kraft paper smears and 29% more alpha actvity. Results show a greater sensitivity with cotton fiber over kraft paper at the 95% confidence level. Regulatory requirements for smear materials are vague. The data demonstrate that the difference in sensitivity of smear materials could lead to a large difference in reported results that are subsequently used for meeting shipping regulations or evaluating workplace contamination levels.

Organic Analyses of Highly Radioactive Solutions Requiring Pretreatment by 137Cs Removal

The beta-emitting radionuclide 137Cs is precipitated from the Savannah River Site (SRS) high-level waste by adding sodium tetraphenylborate (TPB). The concentrated and washed precipitate slurry containing CsTPB is transferred to Defense Waste Processing Facility (DWPF) Salt Processing Cell and hydrolyzed to yield an aqueous product (PHA) free of benzene. The PHA is highly radioactive (1E+11 dpm) and contains some water-soluble organic compounds that need to be analyzed for process control and process history. High performance liquid chromatography (HPLC) methods were developed to analyze these compounds with instruments contained in unshielded fume hoods. Removing 137Cs from PHA in a shielded cell prior to HPLC analyses is essential to minimize personnel radiation exposure. In this study ammonium molybdophosphate (Bio-Rad AMP-1) was used for removing 137Cs in PHA (pH 3.8). With a two-step application of AMP slurry, a decontamination factor of 1E+4 was achieved for Cs removal. Two separate HPLC methods were used to analyze four polar organic compounds. The recovery of all the organic compounds were above 85%.

An algorithm for source checking continuous air monitors using radon progeny.

Department of Energy requirements contained within 10CFR835 require that continuous air monitors be periodically checked for operability. The DOE air monitoring implementation guide for 10CFR835 allows the use of radon progeny to perform the recommended weekly source check. The Defense Waste Processing Facility located at the Savannah River Site has demonstrated that, through the use of the Hypotheses Concerning Two Means, diurnal change in the radon progeny detected by the monitors meets the requirements for weekly source checks. The use of the diurnal change in radon progeny has replaced the person-hours expended performing direct weekly source checking with an automated system requiring minimal person-hour expenditure.

10.2172/918137

The Defense Waste Processing Facility (DWPF) is about to process High Level Waste (HLW) Sludge Batch 4 (SB4). This sludge batch is high in alumina and nepheline can crystallize readily depending on the glass composition. Large concentrations of crystallized nepheline can have an adverse effect on HLW glass durability. Several studies have been performed to study the potential for nepheline formation in SB4. The Phase 3 Nepheline Formation study of SB4 glasses examined sixteen different glasses made with four different frits. Melt rate experiments were performed by the Process Science and Engineering Section (PS&E) of the Savannah River National Laboratory (SRNL) using the four frits from the Phase 3 work, plus additional high B2O3/high Fe2O3 frits. Preliminary results from these tests showed the potential for significant improvements in melt rate for SB4 glasses using a higher B2O3-containing frit, particularly Frit 503. The main objective of this study was to investigate the durability of SB4 glasses produced with a high B2O3 frit likely to be recommended for SB4 processing. In addition, a range of waste loadings (WLs) was selected to continue to assess the effectiveness of a nepheline discriminator in predicting concentrations of nepheline crystallization that would be sufficient to influencemore » the durability response of the glass. Five glasses were selected for this study, covering a WL range of 30 to 50 wt% in 5 wt% increments. The Frit 503 glasses were batched and melted. Specimens of each glass were heat-treated to simulate cooling along the centerline of a DWPF-type canister (ccc) to gauge the effects of thermal history on product performance. Visual observations on both quenched and ccc glasses were documented. A representative sample from each glass was submitted to the SRNL Process Science Analytical Laboratory (PSAL) for chemical analysis to confirm that the as-fabricated glasses corresponded to the defined target compositions. The Product Consistency Test (PCT, ASTM C1285) was performed in triplicate on each Frit 503 quenched and ccc glass to assess chemical durability. The experimental test matrix also included the Environmental Assessment (EA) glass and the Approved Reference Material (ARM-1) glass. Representative samples of all the ccc glasses were examined for homogeneity visually and by X-ray diffraction (XRD) analysis. Chemical composition measurements indicated that the experimental glasses were close to their target compositions. PCT results showed that all of the Fit 503 quenched glasses had an acceptable durability compared to the EA benchmark glass. The durability of one of the ccc glasses, NEPHB-04, was statistically greater than its quenched counterpart. However, this was shown to be of little practical significance, as the durability of the NEPHB-04 ccc glass was acceptable when compared to the durability of the EA benchmark glass. Visual observations and PCT results indicated that all of the Frit 503 quenched glasses were free of any crystallization that impacts durability. For the ccc glasses, XRD results indicated that the lower WL glasses (30 to 40 wt%) were amorphous, which was consistent with visual observations and PCT responses. The higher WL glasses (45 and 50 wt%) were shown by XRD to contain spinel (trevorite, NiFe2O4). It is possible that some of the other high WL glasses also contained some nepheline, but that the amount of nepheline crystallization was below the detection limit (0.5 vol%) associated with XRD. The results indicate that Frit 503 is a good candidate for SB4 processing, based on chemical durability of homogeneous and devitrified glasses over a WL range of 30 - 50%. It should be noted that the higher WL glasses would not be fit for processing in DWPF as they exceed other process related criteria (such as liquidus temperature). However, this is only one of many factors influencing the frit selection. Melt rate and the final SB4 composition are also important factors in frit selection. Additional melt rate studies are currently underway, and the final composition projection for SB4 is expected shortly.« less

The Effect of Adding Crystalline Silicotitanate on the Durability, Liquidus, and Viscosity of Simulated High-Level Waste Glasses at Savannah River Site

This report provides a summary of the results obtained for a limited variability study for glasses containing Crystalline Silicotitanate (CST), Monosodium Titanate (MST), and either simulated Purex or HM sludge. Twenty-two glasses containing Purex sludge and three glasses containing HM sludge were fabricated and tested. The fabricated glasses were tested for durability using the 7-day Product Consistency Test (PCT) and characterized by measuring the viscosity at 1,150 C and by determining an approximate, bounding liquidus temperature. The current models used by Defense Waste Processing Facility (DWPF) for predicting durability, viscosity, and liquidus temperature were applied to all 25 glasses. The goal of this work was to identify any major problems from a glass perspective, within the scope of this effort, which could potentially preclude the use of CST at DWPF.

Characterization of and Waste Acceptance Radionuclides to be Reported for the Second Macro-Batch of High-Level Waste Sludge Being Vitrified in the DWPF Melter

The Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is currently processing the second million gallon batch (Macro-Batch 2) of radioactive sludge slurry into a durable borosilicate glass for permanent geological disposal. To meet the reporting requirements as specified in the Department of Energy's Waste Acceptance Product Specifications (WAPS), for the final glass product, the nonradioactive and radioactive compositions must be provided for a Macro-Batch of material. In order to meet this requirement, sludge slurry samples from Macro-Batch 2 were analyzed in the Shielded Cells Facility of the Savannah River Technology Center (SRTC). This information is used to complete the necessary Production Records at DWPF so that the final glass product, resulting from Macro Batch 2, may be disposed of at a Federal Repository. This paper describes the results obtained from the analyses of the sludge slurry samples taken from Macro-Batch 2 to meet the reporting requirements of the WAPS. Twenty eight elements were identified for the nonradioactive composition and thirty one for the radioactive composition. The reportable radioisotopes range from C-14 to Cm-246.