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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Cesium Isotherm Testing with Spherical Resorcinol-Formaldehyde Resin at High Sodium Concentrations

Washington River Protection Solutions (WRPS) is developing a Low-Activity Waste Pretreatment System (LAWPS) to provide low-activity waste (LAW) directly to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste Facility for immobilization. The pretreatment that will be conducted on tank waste supernate at the LAWPS facility entails filtration to remove entrained solids and cesium (Cs) ion exchange to remove Cs from the product sent to the WTP. Currently, spherical resorcinol-formaldehyde (sRF) resin (Microbeads AS, Skedsmokorset, Norway) is the Cs ion exchange resin of choice. Most work on Cs ion exchange efficacy in Hanford tank waste has been conducted at nominally 5 M sodium (Na). WRPS is examining the possibility of processing supernatant at high Na concentrations—up to 8 M Na—to maximize processing efficiency through the LAWPS. Minimal Cs ion exchange work has been conducted at 6 M and 8 M Na concentrations..

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

Sulfonamide and Sulfonamido-phenol Ligands for Extraction of f-Elements from Alkaline High-Level Waste

Alkaline High Level Waste (HLW) has been accumulated at Hanford and Savannah River Sites as a result of reprocessing for nuclear weapons production during the cold war. A large volume (∼100 MGal) has been accumulated in carbon steel tanks at Savannah River (SRS) and Hanford. The tank waste contains three separate phases generated when NaOH was added to previously acidic Purex raffinates: 1-2 Supernatant liquid, salt-cake, and sludge. The sludge consists mainly of insoluble hydroxides of transition metals while the supernate and salt-cake contain caustic-soluble materials, including salts of highly radioactive fission products Cs(I) and Sr(II). Current treatment of alkaline HLW in SRS includes: i) The Actinide Removal Process (ARP), which is based on sorption of {sup 90}Sr and Actinides (An) on monosodium titanate (MST), also known as 'alpha-strike' process, followed by ii) Caustic Side Solvent Extraction (CSSX),4 which is used for the extraction of {sup 137}Cs by modified calixarenes in a hydrocarbon diluent. Residual actinides in some tanks are removed after CSSX by an additional ARP process commonly referred to as 'alpha-finishing'. Despite the success of ARP for Sr and An removal, as it is a sorption process, it represents the kinetic bottleneck of integrated salt waste processing. Hence potential introduction of additional organic ligands for actinide extraction (in a modified CSSX process) could simplify the overall integrated process, making it more efficient and economical, with less titanate needed and shorter sorption time, as some of the actinide component would be removed during CSSX. In this study tri-sulfonamide and o-sulfonamido-phenol (mono-sulfonamide) ligands have been studied as extractants for Sm(III), which is being used as an Am(III) surrogate. Our prior studies in the group using a tri-sulfonamide (iPr-tsa-B) showed favorable extraction for Sm(III) nitrate salts from alkaline solutions. Mono-sulfonamides possess similar orientation of binding sites to pyrocatechols, which have been found to be good ligands for Am(III) binding and extraction from alkaline media. Tri-sulfonamide of the type iPr-tsa-B6 (1 mM in CH{sub 2}Cl{sub 2} solution) was studied for Ln{sup 3+} extraction using Sm(NO{sub 3}){sub 3}.6H{sub 2}0 (10 and 25 μM) in alkaline solution of NaOH (0.05, 0.1, 0.2, 0.3 mM) / 0.1 M NaNO{sub 3}. Stripping of the organic phase was done using 0.1 M HNO{sub 3} and quantification of Sm{sup 3+} was done using ICP-OES at 359.3 nm. The need to improve stability of the complex led to synthesis of compounds with N-donor site closer to the central benzene ring to facilitate cation-π interactions. Synthesis of tri-sulfonamide type A: a) Chloromethyl methyl ether, SnCl{sub 4}, CH{sub 2}Cl{sub 2}, 0 deg. C, N{sub 2}, 4 h, 57%; b) NaN{sub 3}, reflux in H{sub 2}O/acetone for 22 h, 80%; c) PPh{sub 3}, THF/H{sub 2}O, 22 h, 79%; d) p-toluene sulfonyl chloride, Et{sub 3}N, 1,2-DCE, 22 h. Extraction: Sm(NO{sub 3}){sub 3}.6H{sub 2}O (2 mM) in 5 ml of aqueous NaOH (pH 10.5 - 14) + 6 ml of CH{sub 2}Cl{sub 2} solution of msa (20 mole equiv.) were rotated on a wheel (60 rpm; 20 h). Stripping: 5 ml of 0.1 M HNO{sub 3} + CH{sub 2}Cl{sub 2} solution of msa (after extraction, centrifugation and filtration) was rotated on the wheel (60 rpm; 20 h). Sm{sup 3+} was quantified using UV-Visible spectrophotometry.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Studies, Transport and Treatment Concept for Boilers from Berkeley NPP, England - 13599

In November 2011 Studsvik was awarded a contract to transport five decommissioned boilers from the Berkeley Nuclear Licensed Site in the UK to the Studsvik Nuclear Site in Sweden for metal treatment and recycling. A key objective of the project was to remove the boilers from the site by 31 March 2012 and this was successfully achieved with all boilers off site by 22 March and delivered to Studsvik on 6 April. Four boilers have been processed and the fifth is planned for completion by end of December 2012.The project had many challenges including a very tight timescale and has been successfully delivered to cost and ahead of the baseline programme. This paper describes the project and the experience gained from treatment of the first four boilers. It is the first UK project to send large components overseas for recycling and provides new insight into the processing of Magnox gas-circuit components. (authors)

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

Qualification of Processing Batches for Tank Closure Cesium Removal Through In-Tank Batch Contact Testing - 20470

The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)

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

Experimental and Modeling of Low Pressure Plasma as a Complementary Process for Thermal Waste Treatment - 19270

Polymeric waste contaminated with radionuclides is a challenging wasteform in its raw state. This waste is characterized by its high volume to weight ratio, together with the hazard of its radiolysis and consequently, the generation of flammable and explosive gases (e.g. H2 and light hydrocarbons). Thus, minimizing waste volumes together with stabilizing the wasteform are of high significance in the treatment of contaminated polymers. Thermal decomposition of polymeric waste results in gas products, mainly hydrocarbon molecules (C2-C6), which cannot be released to the atmosphere for environmental reasons. Here, we study a secondary gas treatment reactor, using plasma formed by inductive radio-frequency (13.56 MHz RF), in order to decompose the hydrocarbon gas products into smaller molecules. As a case study, we compare the plasma treatment of gases from two sources: a mixed gaseous product collected from the pyrolytic decomposition of high-density polyethylene (HDPE), and industrial butane gas. We also examine the effects of systematic addition of Ar, N{sub 2} and Ar+N{sub 2} mixtures to the treated gases. In this work, we utilize optical emission spectroscopy (OES) to obtain in situ data from the region in the center of the inductive RF coil. The main emissions from discharges in the gases or their mixtures occur in the 200-700 nm spectral region. Their relative intensities depend strongly on the plasma excitation mode, N{sub 2} flow rate, and gas feed composition. The addition of N{sub 2} to the input gas enhanced hydrocarbon decomposition as compared to the addition of pure Ar. This can be attributed to decomposition by free-radical reactions rather than by particles collisions processes. We performed preliminary modeling of the plasma behavior from ignition to steady state, which takes less than 0.01 s under these conditions. These simulations will help us to better understand and optimize a variety of decomposition reactions. The results of this study suggest that coupling low pressure plasma to conventional thermal decomposition techniques can provide an effective treatment process for contaminated solid plastic waste. (authors)

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

Corrosion Testing of Monofrax K-3 Refractory in Defense Waste Processing Facility (DWPF) Alternate Reductant Feeds

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) uses a combination of reductants and oxidants while converting high level waste (HLW) to a borosilicate waste form. A reducing flowsheet is maintained to retain radionuclides in their reduced oxidation states which promotes their incorporation into borosilicate glass. For the last 20 years of processing, the DWPF has used formic acid as the main reductant and nitric acid as the main oxidant. During reaction in the Chemical Process Cell (CPC), formate and formic acid release measurably significant H 2 gas which requires monitoring of certain vessel’s vapor spaces. A switch to a nitric acid-glycolic acid (NG) flowsheet from the nitric-formic (NF) flowsheet is desired as the NG flowsheet releases considerably less H 2 gas upon decomposition. This would greatly simplify DWPF processing from a safety standpoint as close monitoring of the H 2 gas concentration could become less critical. In terms of the waste glass melter vapor space flammability, the switch from the NF flowsheet to the NG flowsheet showed a reduction of H 2 gas production from the vitrification process as well. Due to the positive impact of the switch to glycolic acid determined on the flammability issues, evaluation of the other impacts of glycolic acid on the facility must be examined.

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

WTP Pretreatment Facility Potential Design Deficiencies--Sliding Bed and Sliding Bed Erosion Assessment

This assessment is based on readily available literature and discusses both Newtonian and non-Newtonian slurries with respect to sliding beds and erosion due to sliding beds. This report does not quantify the size of the sliding beds or erosion rates due to sliding beds, but only assesses if they could be present. This assessment addresses process pipelines in the Pretreatment (PT) facility and the high level waste (HLW) transfer lines leaving the PT facility to the HLW vitrification facility concentrate receipt vessel.

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

The Integrated Waste Feed Delivery Plan for the Direct Feed Low-Activity Waste Mission - 17160

In order to begin tank waste treatment and reduce the risk to the liquid waste cleanup mission, a plan to directly feed the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Vitrification Facility before other WTP facilities has been implemented. Implementation requires another facility, the Low-Activity Waste Pretreatment System (LAWPS), to generate LAW from tank waste supernatant, along with upgrades to the existing tank farm facilities to provide waste feed and receive process returns. The Integrated Waste Feed Delivery Plan (IWFDP) presents the strategy for reliably delivering compliant waste feed to the LAWPS and subsequently WTP to safely and efficiently accomplish the direct feed LAW (DFLAW) mission. Modifications to existing systems and installation of new systems will be coordinated to meet WTP startup, commissioning, and processing needs. The IWFDP presents the process approach, feed campaign plans, and a list of infrastructure projects required to upgrade the DST system in order to support DFLAW start-up and operations. (authors)

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

Robust and High-Waste Loaded Glass Formulations for Hanford Low-Activity Waste - 19050

The Hanford Low-Activity Waste (LAW) vitrification facility will receive LAW batches with abruptly changing compositions that span relatively broad ranges. In particular, the concentration ratios of sodium-to-sulfate, -chloride, -chromate, -phosphate, -potassium, and -aluminate create challenges in glass formulation. Glass compositions that simultaneously satisfy several processability, product quality, and composition constraints with sufficient confidence must be formulated for each LAW batch in near-realtime to enable desired throughput. A method for developing glass compositions while maximizing the waste loading and the process robustness was developed. This method compiled the glass composition-property data collected over nearly two decades, primarily by Pacific Northwest National Laboratory, The Catholic University of America, and Savannah River National Laboratory, in support of both the Hanford Tank Waste Treatment and Immobilization Plant project and the U.S. Department of Energy Office of River Protection Enhanced Waste Glass project. These data were used to develop models that link the melter feed composition to the resulting glass properties. Mathematical constructs were developed to optimize glass compositions for each waste batch in real-time that will simultaneously satisfy processability, product quality, and composition constraints with sufficient statistical confidence. Property prediction and various process uncertainties were reflected in the formulation. The formulation approach directly lends itself to adjustment of glass formulation to achieve program goals. This presentation will summarize the formulation approach and supporting data. (authors)

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

Waste Treatment and Immobilization Plant, a Leader in the International Glass Community - 16003

The U.S. Department of Energy (DOE), Office of River Protection (ORP) is constructing the Tank Waste Treatment and Immobilization Plant (WTP) [1] to treat radioactive waste currently stored in underground tanks at the Hanford Site in Washington State. The WTP that is being designed to separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions with the majority of the mass (approximately 90 percent) directed to LAW and most of the activity (greater than 95 percent) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum bearing solids to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. These two unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A significant consideration to the most recent revision of the glass models is the extensive collaboration developed, funded, managed, and coordinated from the ORP. The current effort includes researchers at the Pacific Northwest National Laboratory, Idaho National Laboratory, Savannah River National Laboratory, Brookhaven National Laboratory, Catholic University of America, Washington State University, Rutgers University, Institute of Chemical Technology Prague, University of Ottawa, Tokyo Institute of Technology, The University of Adelaide, and University of Sheffield. Independent technical evaluation having been solicited from faculty of Alfred University and Vanderbilt University. These advances in glass chemistry generate a need to revisit the feed vectors, flowsheet, and the unit operations driven by minimal waste loading requirements set forth in the contract for the design and construction of the plant. The current status and plans for the immediate future of glass chemistry are the subject of the primary concern to successfully completing our clean-up mission at Hanford. (authors)

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

Ion–ion interactions enhance aluminum solubility in alkaline suspensions of nano-gibbsite (α-Al(OH) 3 ) with sodium nitrite/nitrate

Despite widespread industrial importance, predicting metal solubilities in highly concentrated, multicomponent aqueous solutions is difficult due to poorly understood ion–ion and ion–solvent interactions. Aluminum hydroxide solid phase solubility in concentrated sodium hydroxide (NaOH) solutions is one such case, with major implications for ore refining, as well as processing of radioactive waste stored at U.S. Department of Energy legacy sites, such as the Hanford Site, Washington State. The solubility of gibbsite (α-Al(OH) 3 ) is often not well predicted because other ions affect the activity of hydroxide (OH – ) and aluminate (Al(OH) 4 – ) anions. In the study reported herein, we systematically examined the influence of key anions, nitrite (NO 2 – ) and nitrate (NO 3 – ), as sodium salts on the solubility of α-Al(OH) 3 in NaOH solutions taking care to establish equilibrium from both under- and oversaturation. Rapid equilibration was enabled by use of a highly pure and crystalline synthetic nano-gibbsite of well-defined particle size and shape. Measured dissolved aluminum concentrations were compared with those predicted by an α-Al(OH) 3 solubility model derived for simple Al(OH) 4 – /OH – systems. Specific anion effects were expressed as an enhancement factor (Al enhc ) conveying the excess of dissolved aluminum. At 45 °C, NaNO 2 and NaNO 3 -containing systems exhibited Al enhc values of 2.70 and 1.88, respectively, indicating significant enhancement. The solutions were examined by Raman and high-field 27 Al NMR spectroscopy, indicating specific interactions including Al(OH) 4 – –Na + contact ion pairing and Al(OH) 4 – –NO 2 – /NO 3 – ion–ion interactions. Dynamic evolution of the α-Al(OH) 3 particles including growth and agglomeration was observed revealing the importance of dissolution/reprecipitation in establishing equilibrium. These studies indicate that incomplete ion hydration, as a result of the low water activity in these concentrated electrolytes, results in: (i) enhanced reactivity of the hydroxide ion with respect to α-Al(OH) 3 ; (ii) increased concentrations of Al(OH) 4 – in solution; and (iii) stronger ion–ion interactions that act to stabilize the supersaturated solutions. This information on the mechanisms by which α-Al(OH) 3 becomes supersaturated is essential for more energy-efficient aluminum processing technologies, including the treatment of millions of gallons of Al(OH) 4 – -rich high-level radioactive waste.

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

Molten Salt Oxidation of Ion Exchange and Oils in Carbonate Salts and Evaporator Residue - 16531

Molten Salt Oxidation is a well-studied and robust technology. Recently, renewed interest in the technology has been seen due to limitations on the use of incineration for hazardous and radioactive waste streams. The nuclear research center at Rez, near Prague, has been investigating a bench scale molten salt oxidation system to provide insight for the design of a pilot scale system to condition spent ion exchange resin and scintillator oils. The use of both an eutectic K/Na/Li carbonate salt and a simulated residue salt for liquid waste obtained from evaporated primary coolant of nuclear reactors was examined. In these studies the usable capacity of the salts were examined. Materials doped with surrogate radioisotopes were processed and the capture efficiency was measured. Corrosion of the reactor vessel is of significant engineering concern. Corrosion coupon tests were performed for Inconel HX and stainless steel. These tests provided corrosion rates to assist in design of pilot scale systems. While the bench scale reactor's detention time was too short to provide complete combustion, the values of emissions were tracked and showed an increase in the combustion efficiency at temperatures greater than 750 deg. C. A two part reactor has been designed and work is planned to continue to optimize the system. Additionally, further work has examined the final conditioning of salt which has been saturated with radionuclides after processing organic waste streams A conceptual pilot plant has been designed to solidify radionuclide saturated salt into a form acceptable for repositories according to EU directives. Saturated salt is milled to a powder and immobilized in a matrix with a high percentage of waste loading. Both geo-polymer and polysiloxane matrices have been methods for low temperature immobilization. Conditioning in this manner decreases operating and capital costs when compared to current conditioning technologies and procedures utilized for high volume reduction factors of final waste forms. (authors)

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

Uptake selectivity for different types of zeolites in the presence of boric acid - 15244

The development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima NPP-1. Many removal methods have been developed to separate nuclide from decontaminated solutions containing multi-nuclide so far. From the standpoint of cost efficiency, stability for usage and safety handling, the decontamination methods using inorganic adsorbents are one of the most effective methods for the selective decontamination system. Thus, many researches related to inorganic adsorbents have been carried out after the accident; actually, in Fukushima NPP-1, zeolites are used to remove radioactive nuclides from decontaminated water as a main adsorbent in the circulating injection cooling system. However, the adsorption characteristics of zeolites in the presence of boric acid have not been clarified in detail. As for the decontamination in PWR[1], the adsorption properties should be examined in the presence of highly concentrated boric acid. Based on this background, the final objective of this study is to develop the decontamination methods in PWR-NPP by using various zeolites for multi-nuclide. In this study, the adsorption properties for eight kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Chabazite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ion ({sup 137}Cs{sup +}, {sup 60}Co{sup 2+}, {sup 85}Sr{sup 2+} and {sup 152}Eu{sup 3+}), (2) preparation of Ag-loaded Natural Mordenite (Ag-NM) and uptake properties of I- ions, and (3) multi-nuclide decontamination properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. (authors)

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

Challenges, Successes, and Lessons Learned on Two Active DOE Facility D and D Projects - 18379

'Surgical removal', although having a number of technical and safety challenges, is an effective alternative to traditional Demolition and Decommissioning (D and D) methods. When carefully planned and managed, surgical removal of equipment can be safely performed to achieve cost savings and accelerated schedule. At Fernald, a large groundwater extraction system that treats radionuclides was downsized during continued operations. Challenges included demolition of remedial components in tight spaces adjacent to active components, in-situ management and removal of slurry multimedia and resins, and complex hoisting/rigging of whole vessels from very tight areas. The approach realized a significant reduction in waste volumes and slurry transfer to reduce radionuclide exposure while allowing use of existing vessels as IP-1 shipping containers while providing an opportunity to implement out of the box design for hoisting/rigging/moving large vessels. Several lessons were learned as Los Alamos Technical Associates, Inc. (LATA) faced the project challenges and progressed through the D and D process. The slurry removal required thorough understanding of equipment selection and circulation techniques to maintain a slurry while avoiding plugging of extraction equipment. The physical condition of each individual tank was different and required development of alternate removal techniques, such as draining the vessel and removal of the hard material with an air vacuum truck, for each one. Lessons learned regarding large tank removal focused on recertification of an overhead crane, determination of pick points from a partially filled vessel, reanalyzing the lifting lugs for a partially full tank, and requirements to fill all void space in the vessels to meet off-site disposal specifications. At Oak Ridge National Laboratory (ORNL), LATA performed the surgical removal of all biomass process equipment inside ORNL's Building 2719. Among the challenges was a requirement that the removal action had to be conducted during continued operations. Live steam lines, fire protection, water process and boiler feed water systems were operational during LATA's D and D effort and were in close proximity to removal operations. (authors)

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

Defense Waste Processing Facility Simulant Chemical Processing Cell Studies for Sludge Batch 9

The Savannah River National Laboratory (SRNL) received a technical task request from Defense Waste Processing Facility (DWPF) and Saltstone Engineering to perform simulant tests to support the qualification of Sludge Batch 9 (SB9) and to develop the flowsheet for SB9 in the DWPF. These efforts pertained to the DWPF Chemical Process Cell (CPC). CPC experiments were performed using SB9 simulant (SB9A) to qualify SB9 for sludge-only and coupled processing using the nitric-formic flowsheet in the DWPF. Two simulant batches were prepared, one representing SB8 Tank 40H and another representing SB9 Tank 51H. The simulant used for SB9 qualification testing was prepared by blending the SB8 Tank 40H and SB9 Tank 51H simulants. The blended simulant is referred to as SB9A. Eleven CPC experiments were run with an acid stoichiometry ranging between 105% and 145% of the Koopman minimum acid equation (KMA), which is equivalent to 109.7% and 151.5% of the Hsu minimum acid factor. Three runs were performed in the 1L laboratory scale setup, whereas the remainder were in the 4L laboratory scale setup. Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on nine of the eleven. The other two were SRAT cycles only. One coupled flowsheet and one extended run were performed for SRAT and SME processing. Samples of the condensate, sludge, and off-gas were taken to monitor the chemistry of the CPC experiments.

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

Sludge Washing and Demonstration of the DWPF Nitric/Formic Flowsheet in the SRNL Shielded Cells for Sludge Batch 9 Qualification

Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to qualify the next batch of sludge – Sludge Batch 9 (SB9). Current practice is to prepare sludge batches in Tank 51 by transferring sludge to Tank 51 from other tanks. The sludge is washed and transferred to Tank 40, the current Defense Waste Process Facility (DWPF) feed tank. Prior to sludge transfer from Tank 51 to Tank 40, the Tank 51 sludge must be qualified. SRNL qualifies the sludge in multiple steps. First, a Tank 51 sample is received, then characterized, washed, and again characterized. SRNL then demonstrates the DWPF Chemical Process Cell (CPC) flowsheet with the sludge. The final step of qualification involves chemical durability measurements of glass fabricated in the DWPF CPC demonstrations. In past sludge batches, SRNL had completed the DWPF demonstration with Tank 51 sludge. For SB9, SRNL has been requested to process a blend of Tank 51 and Tank 40 at a targeted ratio of 44% Tank 51 and 56% Tank 40 on an insoluble solids basis.

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

Integration of National Laboratory and Low-Activity Waste Pretreatment System Technology Service Providers - 16435

The National Laboratories are a critical partner and provide expertise in numerous aspects of the successful execution of the Direct-Feed Low Activity Waste Program. The National Laboratories are maturing the technologies of the Low-Activity Waste Pretreatment System consistent with DOE Order 413.3B, 'Program and Project Management for the Acquisition of Capital Assets', expectations. The National Laboratories continue to mature waste forms, i.e., glass and secondary waste grout, for formulations and predictions of long-term performance as inputs to performance assessments. The working processes with the National Laboratories have been developed in procurements, communications, and reporting to support the necessary delivery-based technology support. The relationship continues to evolve from planning and technology development to support of ongoing operations and integration of multiple highly coordinated facilities. (authors)

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

Daher-TLI UCSC Wash Process - Lessons Learned from Handling a Legacy Cylinder - 16506

The Daher-TLI UF{sub 6} Cylinder Service Center (UCSC) began operation in 2013 and functions as a dedicated wash, re-certification, storage and disposal facility for 122 cm uranium hexafluoride (UF{sub 6}) cylinders. The primary purpose of the facility is to provide cylinder cleaning and re--certification services to the UF{sub 6} industry for cylinders that previously contained natural or depleted UF6. The facility also has the ability to recover most of the uranium in the form of uranyl peroxide (UO{sub 4}) or triuranium octoxide (U{sub 3}O{sub 8}). Small residues of UF{sub 6} (commonly referred to as a heel) from natural cylinders are removed and treated in a patented three-step purification process that recovers U{sub 3}O{sub 8} that meets the ASTM specification. The wastewater ordinarily contains a mixture of sodium salts (77% sodium sulfate, 23% sodium nitrate) along with minute traces of uranium (1 - 2 mg/L) and fluoride (10 - 20 mg/L). Over a period of 6 months, lab analyses were performed on the filtrate from a legacy cylinder in an attempt to separate uranium and fluoride from this stream. These pilot results established near quantitative recovery of uranium (< 1 mg/L) and near complete separation of fluorides (5 - 15 mg/L). This paper follows the efforts of the Daher-TLI UCSC to process the abnormal heeled UF{sub 6} cylinder. This paper also documents the lessons learned from this experience and how those lessons were used to improve the UCSC wash process. These modifications served to recover 99.87% of the input uranium as UO{sub 4} while simultaneously creating a waste stream with only trace amounts of impurities. (authors)

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