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At least 469 records · Page 26

Technical Feasibility of Direct Disposal of Electrorefiner Salt Waste

The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of this EBR-II used fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy decided to treat the EBR-II sodium-bonded used fuel in an electrorefiner (ER), which produces a metallic waste, mostly from the cladding. The salt remaining in the ER contains most of the actinides and fission products. Two baseline waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams are being investigated that may reduce the complexity. For example, performance assessments show that both mined repositories in salt and deep boreholes in basement crystalline rock can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Hence the focus of a direct disposal option, as described herein, is now on the feasibility of packaging the ER salt waste in the near term such that it can be transported to a repository in the future without repackaging. A vessel for direct disposal of ER salt waste has been previously proposed, designed, and a prototype manufactured based on desirable features for use in the hot cell. The reported analysis focused on the feasibility of transporting this proposed vessel and whether any issues would suggest that a smaller or larger size is more appropriate. Specifically, three issues are addressed (1) shielding necessary to reduce doses to acceptable levels; (2) the criticality potential and the ease which it can be shown to be inconsequential, and (3) temperatures of the containers in relation to acceptable cask limits. The generally positive results demonstrate that direct disposal of ER in the proposed packaging is feasible without the need to secure funding to modify the facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Consolidation of iodine-loaded Ag 0 -functionalized silica aerogel with HUP, HIP and SPS

The U.S. Department of Energy is currently investigating alternative sorbents for the removal and immobilization of radioiodine from the gas streams in a nuclear fuel reprocessing plant. One of these new sorbents, silver-functionalized silica aerogel (Ag 0 -aerogel), shows promise as a potential replacement for silver mordenite because of its high selectivity and sorption capacity for iodine. Moreover, a feasible consolidation of iodine-loaded Ag 0 -aerogel to a durable SiO 2 -based waste form makes this aerogel an attractive choice for sequestering radioiodine. The main purpose of the study reported here was to investigate consolidation of Ag 0 -aerogel powders with or without iodine and with or without sintering additives with hot uniaxial pressing (HUP), HIP, and SPS. The selected densified samples were analyzed with a helium gas pycnometer for apparent density, with the Archimedes method for open porosity and apparent density, with scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) for the extent of densification, macro- and microstructural changes, and distribution of individual elements, and with X-ray diffraction (XRD) for crystalline phases. To convert the iodine-loaded Ag 0 -silica aerogel to a waste form, rapid consolidation with HUP, HIP, and SPS at moderate temperatures and pressures yields a final waste form consisting of AgI particles encapsulated in fused silica. HUP-processing (1200°C for 30 min under 29 MPa pressure) of Ag 0 -aerogel loaded with 20.2 mass% of iodine provided a product with residual open porosity of 16.9%. However, ~93% of sorbed iodine was retained in the sintered material. Densification with HIP at 1200°C for 30 min under 207 MPa pressure resulted in a fully dense silica-based waste form of 3.3 $\times$ 10 3 kg/m 3 bulk density containing ~22 mass% of iodine. Iodine was retained in the form of nano- and micro-inclusions of AgI dispersed in the silica matrix. Additives were not needed to enhance sintering process as depicted from the lack of open porosity and 100% retention of iodine in the consolidated product. The future tests will tell whether additives such as raw silica aerogel or Ag 0 -aerogel are required for HIPing to produce a fully dense waste form from iodine-loaded Ag 0 -aerogel containing higher concentrations of captured iodine. SPS-processing of samples with or without iodine revealed a benefit of additive-enhanced sintering. The samples containing additives (raw silica aerogel, colloidal silica, and Ag 0 -aerogel) shrunk faster and at lower temperatures. In addition, additives significantly improved distribution of silver compounds in the fully densified products, considerably decreasing the number of micron-sized inclusions of AgI, Ag and Ag 2 S at the grain boundaries.

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Algorithm for Establishing Safeguards Termination Concentration Limits

This paper presents an algorithm for estimating the maximum concentration limits for terminating safeguards on nuclear materials. These limits are important because such material is no longer subject to accounting, reporting or inspections after the International Atomic Energy Agency (IAEA) has approved the termination of safeguards on nuclear material declared by a State to be waste.Maximum concentration limits for different waste forms generated within nuclear fuel cycle facilities must be sufficiently conservative such that termination of safeguards does not become a weak link in the safeguards system. The concentration limits must also remain practical in order to ensure IAEA resources are used effectively and efficiently to implement safeguards in the State. These termination limits should be technically-based while remaining objective and reasonable to implement by the State. In support of the original IAEA guidance on concentration limits for termination of safeguards prepared in the early 1990s, the U.S. developed an algorithm to estimate concentrations that would make recovery of nuclear material from waste on which safeguards had been terminated at least as unattractive as undeclared production from uranium ore or diversion of similar nuclear material. In 2016, the IAEA sought technical advice from a meeting of experts from selected member States to support updating its internal guidance on termination of safeguards. Their recommendations included extending the guidance to consider additional waste forms. The experts in the 2016 meeting recommended that the IAEA limits should more clearly reflect the technical difficulty and level of effort required to recover nuclear material from the various waste forms. This paper identifies and assesses the most likely techniques a State might use to recover nuclear material from the waste forms identified by the experts and determines how the safeguards termination algorithm could be revised to more accurately reflect the difficulty of recovering one significant quantity of nuclear material from these waste forms.

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Update to the Performance Assessment for the Savannah River Site Saltstone Disposal Facility - 20124

In 2019, Savannah River Remediation developed a revision to the performance assessment (PA) on behalf of the U.S. Department of Energy (DOE) Savannah River Operations Office (SR) for the near-surface disposal of low-level waste at the Savannah River Site (SRS) Saltstone Disposal Facility (SDF). Soluble waste from SRS Tank Farms undergoes salt processing to remove cesium and other high activity constituents. The low-activity decontaminated salt solution (DSS) is then immobilized by mixing it into a cementitious waste form known as saltstone. After mixing, the saltstone is poured into leak-tight concrete vaults, known as saltstone disposal units (SDUs), where the waste form cures. By the time of facility closure, the SDF is expected to consist of 15 SDUs with a combined capacity of 1.06 E+09 L (280 Mgal) of cured saltstone. The facility operates under a Disposal Authorization Statement (DAS) from DOE and a permit from the South Carolina Department of Health and Environmental Control (SCDHEC). Since the start of operations in 1990, the SDF has received almost 6.7 E+07 L (18 Mgal) of DSS, resulting in the safe disposal of 2.7 E+16 Bq (7.3 E+05 Ci) of activity. Due to the radioactive decay of short-lived contaminants, the total remaining activity in the disposed waste is estimated to be approximately 1.4 E+16 Bq (3.9 E+05 Ci), as of September 2018. The Disposal Authorization Statement requires a demonstration that the system of engineered and natural features of the disposal facility will limit releases from the facility and be protective of human health and the environment for at least the next 1,000 years. The long-term performance of the facility was evaluated under the requirements of the DoE's Radioactive Waste Management Manual (US DOE Manual 435.1-1). Simulations were performed to demonstrate that the disposal facility would meet performance objectives specified in the manual. The evaluation was based on numerical models that simulate the releases of contaminants from the saltstone waste form. Contaminants were transported through groundwater and air pathways to points of assessment to evaluate compliance. In addition, the potential consequences of an inadvertent human intrusion (IHI) were also evaluated. In accordance with the guidance and recommendations in US DoE's technical standard for DAS, deterministic and probabilistic analyses were performed to demonstrate the SDF system, which includes the engineered cover system, the SDUs, the waste form, and the natural features of the site, provides a reasonable expectation that saltstone disposal will meet performance objectives. Post-closure doses to future members of the public (MOP) were evaluated; dose estimates from the air pathway were well below the 1.0 E-04 Sv/yr (10 mrem/yr) performance objective within 1,000 years, and the calculated doses to the MOP from all exposure pathways, including the groundwater and air pathways, were well below the 2.5 E-04 Sv/yr (25 mrem/yr) performance objective within 1,000 years. Doses following an assumed intrusion event within the facility boundaries were well below the IHI performance objectives, where the acute IHI dose was below the acute IHI dose performance objective of 5.0 E-03 Sv (500 mrem) and the chronic IHI dose was below the chronic IHI dose performance objective of 1.0 E-03 Sv/yr (100 mrem/yr) within 1,000 years. (author)

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Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization]

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

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Handling and Emplacement Options for Deep Borehole Disposal Conceptual Design

This report presents conceptual design information for a system to handle and emplace packages containing radioactive waste, in boreholes 16,400 ft deep or possibly deeper. Its intended use is for a design selection study that compares the costs and risks associated with two emplacement methods: drill-string and wireline emplacement. The deep borehole disposal (DBD) concept calls for siting a borehole (or array of boreholes) that penetrate crystalline basement rock to a depth below surface of about 16,400 ft (5 km). Waste packages would be emplaced in the lower 6,560 ft (2 km) of the borehole, with sealing of appropriate portions of the upper 9,840 ft (3 km). A deep borehole field test (DBFT) is planned to test and refine the DBD concept. The DBFT is a scientific and engineering experiment, conducted at full-scale, in-situ, without radioactive waste. Waste handling operations are conceptualized to begin with the onsite receipt of a purpose-built Type B shipping cask, that contains a waste package. Emplacement operations begin when the cask is upended over the borehole, locked to a receiving flange or collar. The scope of emplacement includes activities to lower waste packages to total depth, and to retrieve them back to the surface when necessary for any reason. This report describes three concepts for the handling and emplacement of the waste packages: 1) a concept proposed by Woodward-Clyde Consultants in 1983; 2) an updated version of the 1983 concept developed for the DBFT; and 3) a new concept in which individual waste packages would be lowered to depth using a wireline. The systems described here could be adapted to different waste forms, but for design of waste packaging, handling, and emplacement systems the reference waste forms are DOE-owned high- level waste including Cs/Sr capsules and bulk granular HLW from fuel processing.

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Leaning the tank closure process at the Savannah River Site - 15515

The Savannah River Site liquid waste system consists of multiple facilities that safely receive and store legacy radioactive waste prior to treatment and disposition. The liquid radioactive waste is stored in large underground tanks. The waste inventory currently stored in these tanks is a complex mixture of insoluble metal hydroxide solids and soluble salt supernate. The high-level (radioactive) portion of the waste is vitrified into a glass waste form, while the low-level waste is immobilized into a cementitious grout waste form called salt-stone. Once the waste is retrieved and processed, the tanks are closed. The closure process includes; waste removal, characterizing residuals, isolating the tank systems, and stabilizing the tank with tailored grout formulations. The core of Lean is to eliminate waste or non-value added steps from processes to maximize customer value. From the perspective of the customer, 'value' is defined as any action or process for which a customer would willingly pay. In this case our product is to safely and efficiently remove waste from the underground storage tanks and demonstrate operational closure to the State of South Carolina. Savannah River Remediation (SRR) initiated the use of the Lean process in late 2013 and identified the Tank Closure process as one of the targeted areas to improve. The first step in the Lean process is to 'map' out the value stream and identify areas where non-value added steps are present and then prioritize Rapid Improvement Events (RIE). Based on this mapping process, there were a number of areas that SRR determined warranted further evaluation for improvements, including; standard equipment design, characterization, standard work packages, regulatory documentation, tank characterization, structural evaluations, engineering documentation and equipment re-use. This paper focuses on the application of the Lean process at SRR and more specifically, effective application to the Tank Closure Process. This paper discusses the value stream mapping process for Tank Closures, as well as, summarizes the RIEs and demonstrate the realized benefits. Benefits can come in a variety of forms, from schedule compression to decreased costs. Through the Lean process we have been able to identify close to $10 M in savings over the life cycle of the tank closure process and a 4 year schedule compression. Cost and schedule savings are used to accomplish further risk reduction efforts. In addition to hard benefits, there are also soft benefits. These soft benefits are a result of the process itself and help drive improved efficiencies through better communication with key stakeholders, such as regulators and customers. (authors)

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Characterization of Technetium Speciation in Cast Stone

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for the U.S. Department of Energy (DOE) EM-31 Support Program (EMSP) subtask, “Production and Long-Term Performance of Low Temperature Waste Forms” to provide additional information on technetium (Tc) speciation characterization in the Cast Stone waste form. To support the use of Cast Stone as an alternative to vitrification for solidifying low-activity waste (LAW) and as the current baseline waste form for secondary waste streams at the Hanford Site, additional understanding of Tc speciation in Cast Stone is needed to predict the long-term Tc leachability from Cast Stone and to meet the regulatory disposal-facility performance requirements for the Integrated Disposal Facility (IDF). Characterizations of the Tc speciation within the Cast Stone after leaching under various conditions provide insights into how the Tc is retained and released. The data generated by the laboratory tests described in this report provide both empirical and more scientific information to increase our understanding of Tc speciation in Cast Stone and its release mechanism under relevant leaching processes for the purpose of filling data gaps and to support the long-term risk and performance assessments of Cast Stone in the IDF at the Hanford Site.

Cast Stone↗

Roll Compaction of Powdered Radioactive Wastes - 18251

Most of low- and intermediate-level radioactive wastes must be solidified by a proper binding matrix and disposed at the centralized disposal site. So it is desirable to achieve a high loading of radwastes in a drum to reduce the disposal fee while meeting both the national regulations and the waste acceptance criteria of the disposal site. In general, the radioactive wastes (liquid, dried powder and sludge) were solidified by mixing with a proper binding matrix (cement, polymer, asphalt, etc.), but their solidified volume inevitably increases due to the limitation of incorporating of radwaste into a binding matrix. To solve this problem, some Korean NPPs operate both a CTS (Concentrate Treatment System, evaporation and granulation) and a polymer solidification system. The size distribution of granulated wastes produced by CTS is so broad (diameter; several μm ∼ several cm) that a polymer cannot impregnate the void space between particles due to their small size. However, if the powdered radwaste can be compacted into high density pellets, and the pellets can be loaded in a drum with maximum waste loading, then the binding matrix can fill the void between pellets, it might be more desirable to keep the size distribution of pellets constant to reduce the solidified waste volume. This compaction methodology can greatly reduce the disposal fee due to the reduction of the solidified waste volume, and can facilitate polymer solidification due to the constant size of the radwaste pellets. This methodology is also applicable to solidify powdered wastes such a depleted uranium used as a catalyst, ash and sludge. Loading radwaste compacted with bentonite pellets versus particles into a container serves, as an engineered or natural barrier in a deep geologic disposal system. Roll compaction is a typical agglomeration process for dry powders and depends on the device parameters and the material properties. This study, firstly, was focused on the design of roll compactor (briquetting machine), the shape and size of pellet, and secondly the loading method of pellets into a drum, their solidification in polymer, and evaluating the overall volume reduction. We used bentonite (particle size; 85 ∼ 100 μm) as a material to be compacted instead of radwaste powders (concentrates and depleted uranium wastes) without the addition of agglomeration agents. In the design of roll, we considered the diameter (200 mm) and width (41 mm) of roll, two types of pocket shapes (< 9 mm, 0.275 ml), and the surface of the pocket. Fig. 2 shows two pellet shapes, tetragonal and circular. The pressure of roll compactor, one of the main process variables, is > 320 kg f/cm{sup 2}, and the roll gap is 0.5 mm. The powders are compacted in a compaction zone between two counter-rotating rolls according to the complicated mechanisms. As a result of the state and the apparent density of the pellets obtained by changing the pressure from 0 to 286 kg f/cm{sup 2}, the pellets were unstable at pressure=0, but above 148 kg f/cm{sup 2} the pellets were more stable, and the apparent density of the pellets was 2.63 g/cm{sup 3} at 148 kg f/cm{sup 3}, and 2.68 g/cm{sup 3} at 286 kg f/cm{sup 2}. The roller speed (rpm) can determine the capacity (kg/h) of a compactor, and depends on both the compaction pressure and the feeding rate. By increasing the roll speed from 2 to 4 rpm, the apparent density of pellets was decreased gradually from 2.74 g/cm{sup 3} to 1.45 g/cm{sup 3}. We designed the special feeding system in order to feed powders uniformly to the roll gripping zone and to supply powders into the narrowest part of the roll gap. At a constant roll speed (3 rpm) and a constant compaction pressure (286 kg f/cm{sup 2}), the apparent density of the pellets was linearly proportional to the feeding rate from 2.18 to 2.74 g/cm{sup 3} by changing the feeding rate of powders from 8 to 12 rpm. In the polymer solidification of the obtained pellets, we used epoxy resin as a solidification agent, and showed their solidified waste forms in Fig. 5. The incorporation ratio (wt.%) of powders/polymer was about 71.7/28.3 in tetragonal pellets, 71.9/28.1 in circular pellets. We calculated the volume reduction in an aspect of powders itself and the solidified waste form. In powder case, through the comparison of the apparent density of the obtained pellets (2.74 g/cm{sup 3}) and raw powders (0.93 g/cm{sup 3}), the volume reduction was calculated as about 2.74/0.93 = 2.95 times. This means that powders can be compacted to 1/2.95 times. And in waste form case, the weight ratio of powder/polymer in a mixing method (workable) and a compaction method is about 0.42, 2.56, individually. Therefore the volume reduction was calculated as about 2.56/0.42 = 6.01 times. We found a compaction methodology for the volume reduction of powdered radioactive wastes, and their polymer solidification will be more effective to reduce the volume of radwastes to be disposed. After a characterization of the solidified waste form incorporated pellets (including sintered pellets), this methodology will be applied to treat depleted uranium wastes used as catalyst, and to solve the problem of NPPs' polymer solidification system, as mentioned above. (authors)

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Engineering scale demonstration of a prospective cast Stone process - 15584

A cementitious waste form is one of the alternatives being evaluated for Supplemental Immobilization of Hanford Low Activity Waste (LAW). Washington River Protection Solutions (WRPS) is collecting data on cementitious or pozzolanic waste forms such as Cast Stone. The goal of this project for WRPS is to obtain data on the performance of the Cast Stone waste form for immobilizing LAW. As part of the data package being assembled, an engineering-scale demonstration with non-radioactive simulants was performed at the Savannah River National Laboratory (SRNL) using the Scaled Continuous Processing Facility (SCPF) to fill a container with simulated Cast Stone grout. Legacy salt solution from previous Hanford salt waste testing was trimmed to resemble the average composition generated from the Hanford Tank Waste Operation Simulator (HTWOS) used in the screening tests. The dry blend materials, ordinary portland cement (OPC), Class F fly ash, and ground granulated blast furnace slag (GGBFS or BFS) were obtained from Lafarge North America in Pasco, WA. Over three days, the SCPF was used to fill a 6,056 liter (1600 gallon) container with simulated Cast Stone grout. The tank was instrumented with x-, y-, and z-axis thermocouples to monitor curing temperature and two formed core sampling vials. The target production rate was 5.7 L/min (1.5 gpm). This required a salt solution flow rate of approximately 3.78 L/min (1 gpm) and a premix feed rate of approximately 263 kh/h (580 lb/h). The final surface slope at a fill height of 100 cm was 2.5-3.8 cm across the 260 cm (8.5 ft) diameter tank. During processing, grout was collected from both the mixer discharge and the discharge into the tank. These samples were stored in a humid environment either in a closed box proximal to the tank or inside the laboratory. (authors)

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Competitive TcO4-, IO3-, and CrO42- Incorporation into Ettringite

Ettringite is a naturally occurring mineral found in cementitious matrices that is known for its ability to incorporate environmentally mobile oxyanion contaminants. To better assess this immobilization mechanism for contaminants within cementitious waste forms intended for nuclear waste storage, this work explores how mixed oxyanion contaminants compete for ettringite incorporation and influence the evolving mineralogy. Ettringite was precipitated in the presence of TcO4-, IO3-, and/or CrO42-, known contaminants of concern to nuclear waste treatment, over pre-determined precipitation periods. Solution analyses quantified contaminant removal and the collected solid was characterized using bulk and microprobe XRD coupled with PDF and microprobe XRF analyses. Results suggest that =96% IO3- is removed from solution, regardless of ettringite precipitation time or the presence of TcO4- or CrO42-. However, TcO4- removal remained <20%, was not significantly improved with longer ettringite precipitation times, and significantly decreased in the presence of IO3-. When IO3- is comingled with CrO42, gypsum is formed as a secondary mineral phase, which results in oxyanion partitioning among mineral phases, e.g., IO3- incorporation into ettringite and CrO42- incorporation into gypsum via SO42- substitution. Results from this work exemplify the importance of competitive immobilization when assessing waste form performance and environmental risk of contaminant release.

Gillispie, Elizabeth C.↗

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

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Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

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Large Diameter Deep Borehole (LDDB) Disposal Design Option for Vitrified High-Level Waste (HLW) and Granular Wastes

This study considers the feasibility of large diameter deep boreholes for waste disposal. The conceptual approach considers examples of deep large diameter boreholes that have been successfully drilled, and also other deep borehole designs proposed in the literature. The objective for large diameter boreholes would be disposal of waste packages with diameters of 22 to 29 inches, which could enable disposal of waste forms such as existing vitrified high level waste. A large-diameter deep borehole design option would also be amenable to other waste forms including calcine waste, treated Na-bonded and Na-bearing waste, and Cs and Sr capsules.

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SaltStone Wastewater Cement Study Using Isothermal Calorimetry, Standard Concrete Characterization Techniques, and CemGEMS - 25169

Cementitious reagents are used to solidify/stabilize aqueous radioactive, hazardous, and mixed salt solutions, and sludges to meet low-level radioactive waste (LLW) and Resource Conservation and Recovery Act (RCRA) requirements for disposal at Department of Energy (DOE). It is flexible enough to solidify radioactive wastewater saturated in complex species that include but are not limited to radioactive isotopes from the bombardment of neutrons in reactor operation, corrosion products from metallic components, and a variety of soluble organic compounds. [1,2]. Waste form testing typically includes processing or fresh properties, cured properties, compressive strength and hydraulic properties, porosity, density, saturated and unsaturated moisture transport, and leachability of contaminants in the waste form pore solution. Properties are collected over a relatively limited time, typically 28 to 365 days [3]. In addition, changes in the waste form as the result of time and changing conditions are important for concrete engineers to predict overall performance of the forms and potential release of contaminants in the disposal process via unintended filtration into the environment [4]. These predictions are determined/calculated characterizing young waste forms (relative to the standard age of concrete) and are based on transport through soluble ions in pore solutions. Characterization methods include X-ray, SEM, and isothermal calorimetry among other methods used to define the composition, amorphous vs. crystalline nature of the components, and the energetic formation mechanisms for multi-phase mineral systems. [5–7] Isothermal calorimetry is a well standardized technique for cements and concretes and can be used to predict the timing and nature of the hydration reactions.[8] The technique can measure long term energetic

Bustamante, Michael E. [Savannah River National La↗

Immobilization in Cement of Rice Husk used as Bio-sorbent of Radionuclides Present in Radioactive Liquid Waste - 17225

The use of new techniques for the treatment of radioactive liquid wastes has been studied in the last few years by many countries. In this context, the bio-sorption has been an attractive process. In Brazil the Radioactive Waste Management Laboratory of Nuclear and Energy Research Institute (IPEN-CNEN/SP), studied some bio-masses to be used as bio-sorbents to remove radionuclides of the radioactive liquid waste stored in this institute. The rice husk is a low cost and wide available byproduct of the agriculture and its bio-sorption capacity has been proved as feasible. Thus, it is necessary to evaluate if its immobilization in a cementitious matrix will result in a waste form that meets the requirements of Brazilian regulations for final disposal of low and intermediate level waste. This work aimed at evaluating the properties of the immobilization of rice husk in a cementitious matrix. Some properties of the waste form with incorporated rice husk were evaluated, as free water, workability, setting time and mechanical strength. The results indicated that cement matrices with 5 and 10% of biomass and water/cement ratio of 0.30 and 0.35 are viable for the production of waste forms that meets the Brazilian regulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A Case for Molecular Recognition in Nuclear Separations: Sulfate Separation from Nuclear Wastes

In this paper, we present the case for molecular-recognition approaches for sulfate removal from radioactive wastes via the use of anion-sequestering systems selective for sulfate, using either liquid liquid extraction or crystallization. Potential benefits of removing sulfate from the waste include improved vitrification of the waste, reduced waste-form volume, and higher waste-form performance, all of which lead to potential cleanup schedule acceleration and cost savings. The need for sulfate removal from radioactive waste, especially legacy tank wastes stored at the Hanford site, is reviewed in detail and primarily relates to the low solubility of sulfate in borosilicate glass. Traditional methods applicable to the separation of sulfate from radioactive wastes are also reviewed, with the finding that currently no technology has been identified and successfully demonstrated to meet this need. Fundamental research in the authors laboratories targeting sulfate as an important representative of the class of oxoanions is based on the hypothesis that designed receptors may provide the needed ability to recognize sulfate under highly competitive conditions, in particular where the nitrate anion concentration is high. Receptors that have been shown to have promising affinity for sulfate, either in extraction or in crystallization experiments, include hexaurea tripods, tetraamide macrocycles, cyclo[8]pyrroles, calixpyrroles, and self-assembled urea-lined cages. Good sulfate selectivity observed in the laboratory provides experimental support for the proposed molecular-recognition approach.

Moyer, Bruce A↗