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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Effect of Calcination Temperature on the Electrochemical Properties of Nickel-rich LiNi0.76Mn0.14Co0.10O2 Cathodes for Lithium-ion Batteries

High energy density nickel (Ni)-rich layered LiNixMnyCozO2 (NMC, x = 0.6) materials are promising cathodes for lithium-ion batteries. However, several technical challenges, such as fast capacity fading, are hampering their large scale application. Herein, we demonstrate that a Ni-rich cathode LiNi0.76Mn0.14Co0.10O2 (NMC76) calcined at an optimal temperature of 750~775oC exhibits a high discharge capacity (~215 mAh g-1 when charged to 4.5 V) and an excellent high rate capability. Even at a 10C rate, it still exhibits a capacity of more than 160 mAh g-1 and retains ca. 79% of its initial capacity after 200 cycles. The high performance of NMC76 is directly related to its optimized primary particle size (that constitute the spherical secondary particles of >10 m) and cation mixing. Higher calcination temperature (= 800 oC) leads to rapid increase of primary particle size, poor cycling stability, and inferior rate capability of NMC76 due to its severe micro strain and crack formation upon repeated lithium ion de/intercalations. Therefore, NMC76 calcined at 750~775oC is a very good candidate for the next generation of Li ion batteries.

nickel-rich cathodes, calcination temperature, cyc↗
DOE OSTI2019

Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

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

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

This comprehensive report provides definitive information regarding the volume, mass, and composition (chemical and radioactivity) of the calcined waste stored at the Idaho Nuclear Technology and Engineering Center on the Idaho National Laboratory Site. The calcine composition data are required for current regulatory compliance and safe operation of the INTEC’s six Calcined Solids Storage Facilities (CSSFs), as well as future retrieval, treatment, and shipping the calcine to an off-Site facility (such as a geologic repository). This report also contains a description of the calcine storage bins that compose each of the CSSFs, emphasizing the features that may be of concern to future calcine retrieval. The CSSFs were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF designs will affect future calcine retrieval processes and equipment. This revision (5) of the report presents refinements of the databases used to calculate the volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. This revision corrects minor errors and inconsistencies noted in an extensive validation and verification effort of the prior version of the Microsoft Excel workbooks upon which the calcine composition described in this report is based. Historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all of the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, the volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2010) databases and spreadsheets. The Excel spreadsheets are collectively called the Historical Processing Model (HPM). The calcine composition determined by this method compares favorably with historical calcine sample data.

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

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

Vitrification Process for Idaho HLW Calcine - 18433

The majority of the liquid wastes produced from reprocessing of several types of spent nuclear fuel at the Department of energy's (DOE) Idaho site have been volume reduced by calcination. The resulting ∼155,600 ft{sup 3} of calcined waste is stored in 43 shielded stainless steel bins, which have a 500-year design life. The calcined material is both a high-level waste and a RCRA hazardous waste, which complicates treatment and disposal options. In addition to a potential direct disposal option, treatment options that have been considered include steam reforming, hot isostatic pressing (HIP), and vitrification. This paper describes testing and evaluation of a vitrification option for the calcine. Vitrification to produce a borosilicate glass has the advantage that it is the only treatment that provides a waste form that meets the current HLW repository waste acceptance criteria. It is therefore the lowest risk option in terms of regulatory and legislative concerns. The present work addressed several key questions: (i) Calcine waste loadings in glass, which determine the number of HLW canisters produced and the required glass production rate (throughput) for the system design; (ii) Melt rate per unit glass pool surface area of the melter, which determines the sizing of the vitrification system; and (iii) The ability to meet waste form acceptance requirements (PCT and TCLP) while achieving reasonable waste loadings. These are issues of fundamental importance to the sizing estimates for the unit operations and overall flow-sheet and they, therefore, directly impact the cost and schedule estimates. Fully compliant, high-waste-loading glasses were developed for two of the four Idaho calcine wastes. The results demonstrate that waste loadings of 35 wt% waste oxides are achievable for the two alumina calcines and about 30 wt% is achievable for the two zirconia calcines. Testing of these formulations in a joule-heated ceramic melter showed glass production rates of over 5000 kg/(m2/d) for the Alumina calcine and over 8500 kg/(m{sup 2}/d) for the Zirconium-Na Blend calcine. These extremely high production rates translate directly into smaller system size and shorter mission duration and, therefore, reduced costs. Based on these results, a preliminary design for the full-scale vitrification system was developed. The system consists of waste receipt, vitrification, off-gas treatment, and canister sealing and decontamination equipment. The calcine is blended with the appropriate glass forming chemicals (or glass frit) and fed to the melter as a solid stream. The melter has a glass pool surface area of about 2.5 m{sup 2} and is equipped with a bubbling system to increase processing rates. Glass production rates of about 6 metric tons per day are projected. The vitrification system would be installed into the existing Idaho Waste Treatment Unit Facility (IWTU) so that no new facilities are required to house the treatment system, thus greatly reducing overall cost, duration, and risk. The system would be designed to treat all of the calcine waste within a period of 12 years at an operating efficiency of 80%, producing a total of about 8500 canisters each weighing two metric tons. (authors)

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

Role of calcination atmosphere in vanadium doped Li{sub 4}Ti{sub 5}O{sub 12} for lithium ion battery anode material

Highlights: • First comparison on V doped Li{sub 4}Ti{sub 5}O{sub 12} (VLTO) synthesized in air and argon. • Ti{sup 4+} do not reduce to Ti{sup 3+} in air calcined VLTO. • V doping enhances reducing of Ti{sup 4+} to Ti{sup 3+} in argon calcined VLTO. • Li{sub 4}Ti{sub 4.9}V{sub 0.1}O{sub 12} calcined in argon shows best charge-discharge behaviors. • Conductivity enhancement in argon calcined VLTO upto three orders. - Abstract: Li{sub 4}Ti{sub 5}O{sub 12} (LTO) with enhanced properties can replace the conventional carbonaceous anode material of lithium ion battery. Vanadium doped LTO (Li{sub 4}Ti{sub 5-x}V{sub x}O{sub 12}, x = 0, 0.05, 0.1, 0.15) materials are synthesized by sol-gel process followed by calcination of dried gel in air and argon atmosphere. No additional phase corresponding V is indicated in XRD patterns. Calcination in argon reduces the particle size of Li{sub 4}Ti{sub 5-x}V{sub x}O{sub 12} while V doping reduces particle size distribution. XPS reveals that Ti{sup 4+} get reduced to Ti{sup 3+} only in argon calcined Li{sub 4}Ti{sub 5-x}V{sub x}O{sub 12} which is enhanced by V doping. Electrochemical impedance spectroscopy confirms that compared to LTO, lithium ion diffusion coefficient of air and argon calcined Li{sub 4}Ti{sub 4.85}V{sub 0.15}O{sub 12} is approximately two and three times higher. Argon calcined Li{sub 4}Ti{sub 4.9}V{sub 0.1}O{sub 12} shows highest charge-discharge capacities at almost all C-rates (0.1 to 5C; 1C = 175mAg{sup −1}). Electronic conductivity of argon calcined Li{sub 4}Ti{sub 4.9}V{sub 0.1}O{sub 12} is 1.7 × 10{sup −2} S cm{sup −1} which is three orders higher than that of LTO.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Kinetic assessment of pulp mill-derived lime mud calcination in high CO 2 atmosphere

The chemical pulping of biomass involves the recycling of calcium through the calcination of lime mud, which is mostly comprised of calcium carbonate (CaCO 3 ). Lime mud decomposes under elevated temperatures to generate calcium oxide (CaO) and carbon dioxide (CO 2 ), the kinetics of which are strongly influenced by the CO 2 partial pressure and temperature. Oxy-fuel combustion and electrified lime kilns for lime mud calcination are intriguing methods to decarbonize this highly polluting operation within the biomass pulping industry. However, the high CO 2 concentration in oxy-fuel and electrified calcination processes alters the kinetics and overall reactivity of lime mud. For the first time, a model-fitting method is used to determine the kinetic parameters for lime mud calcination under a wide range of temperatures (550 °C–1250 °C) and under different concentrations of CO2 (0 %, 15 %, 50 %, and 90 %). A kinetic model is developed that accurately predicts the reaction rates as a function of temperature and CO 2 concentration. The apparent activation of energy for lime mud calcination is elevated under a high CO 2 environment. Relative to inert gas (N 2 , Ar), the temperature window for calcination is much smaller under high CO 2 environments. The presence of Na in lime mud does not seem to affect calcination under a high CO 2 environment. Finally, particle size variation does not have a significant effect on calcination under a high CO 2 environment.

09 BIOMASS FUELS↗
DOE OSTI2016

To study the effect of dopant NiO concentration and duration of calcinations on structural and optical properties of MgO-NiO nanocomposites

In present work Magnesium oxide (MgO) samples were doped with different concentration of Transition metal Nickel Oxide(NiO) by using Chemical co-precipitation method. The doping levels were varied from NiO (5%, 10%, 15%) and all the samples were calcined at 600°C for 4hrs and 8hrs respectively. Structural analysis of these calcined materials is carried out by X-ray diffraction (XRD) techniques which reveals that average crystalline sizes are in nano region i.e. 21.77nm-31.13 nm and tabulated in table 1. The powder of calcined samples were also characterized by using various other techniques i.e. Scanning Electron Microscopy (SEM), Fourier Transformation Infrared Spectroscopy (FTIR), UV-Visible spectroscopy, Transmission Electron Microscopy (TEM) etc. The effects of dopant concentration, calcined temperature, calcinations duration on samples were studied and also investigate the effect of varying dopant concentration on morphology and optical properties of calcined nanomaterials. From results it was observed that the crystallite size of nanocomposites increases with increases dopant concentration or increases calcinations duration. The optical band gap decreases with increases sintering time and increase with increases dopant concentrations. TEM results coincide with XRD results and show that particles are polycrystalline in nature. FTIR spectra show that for all samples particles are pure in composition and transmission rate increases with calcinations duration.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Characterization and recovery of rare earth elements and other critical metals (Co, Cr, Li, Mn, Sr, and V) from the calcination products of a coal refuse sample

Here, the impacts of calcination on the leaching recovery of rare earth elements (REEs) and other critical metals from a Baker seam coal refuse material were investigated. Acid leaching test results showed that leaching recovery of the REEs, especially the light REEs, was increased significantly from around 24% to 80–90% after calcination under 600–750 °C for two hours. By considering the major element leaching kinetics and the mineralogical changes of the dominant minerals caused by calcination, two mechanisms were proposed for the improved leaching performance: (1) decomposition of difficult-to-dissolve REE-bearing minerals into soluble forms; and (2) liberation of REE-bearing minerals encapsulated in clays after calcination due to the dehydration and disintegration of the layered clay structure. Sequential chemical extraction tests showed that a significant fraction (around 75%) of the REEs in the raw material were present in an insoluble form. After calcination under proper temperatures (600–800 °C), the majority of the insoluble light REEs (LREEs) were converted into soluble forms. Changes in the modes of occurrence of the LREEs and phosphorous based on leaching rate and sequential leaching correlated well as a function of the calcination temperature. This finding in addition to SEM-EDX results tends to indicate the potential that the LREEs occur primarily as rare earth phosphates such as monazite in the untreated coal refuse material. The acid leaching and sequential chemical extraction results show that co-extraction of REEs and other critical metals such as Co, Mn, and Li is possible using a calcination-acid leaching procedure.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2015

Comparative study of phase transition and textural changes upon calcination of two commercial titania samples: A pure anatase and a mixed anatase-rutile

The effect of calcination temperature and time on structural and textural changes of two commercial TiO{sub 2} samples (pure anatase and a mixture of anatase and rutile) has been investigated using N{sub 2} physisorption, ex-situ and in-situ X-ray powder diffraction, differential scanning calorimetry and UV–vis diffuse reflectance spectroscopy. The increase of the calcination temperature (up to 700 °C) and time (up to 8 h) causes only textural changes in the pure anatase, whereas a transformation of the anatase to rutile takes place, in addition, in the mixed titania (containing anatase and rutile). The textural changes observed in pure anatase sample were attributed to solid state diffusion leading to an increase in the size of anatase nanocrystals, through sintering. Thus, the mean pore diameter shifts to higher values and the pore volume and specific surface area decrease. The successful application of the Johnson–Mehl–Avrami–Kolmogorov model in the kinetic data concerning the pure anatase indicates a mass transfer control of sintering process. Similar textural changes were also observed upon calcination of the sample containing anatase and rutile. In this case not only sintering but the anatase to rutile transformation contributes also to the textural changes. Kinetic analysis showed that the rutile nanocrystals in the mixed titania served as seed for by-passing the high energy barrier nucleation step allowing/facilitating thus the anatase to rutile transformation. A fine control of the anatase to rutile ratio and thus of energy-gap and the population of hetero-junctions may be obtained by adjusting the calcination temperature and time. - Graphical Abstract: Dependence of anatase content of P25 on the calcination temperature (600 °C (■), 650 °C (●), 700 °C (▲)) and time. - Highlights: • Increase of calcination temperature up to 800 °C and time up to 8 h causes only textural changes in pure anatase • Progressive transformation of anatase to rutile with time takes place in the mixed titania above 600 °C • A high activation energy barrier inhibits the solid state transformation in pure anatase • Rutile nanocrystals in mixed titania serve as seeding for favouring transformation • Calcination temperature and time allow a fine control of E{sub g} and heterojunctions population in mixed titania.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Summary of Calcine Disposal Development Using Hot Isostatic Pressing

Battelle Energy Alliance, LLC, has demonstrated the effectiveness of the hot isostatic press (HIP) process for treatment of hazardous high-level waste known as calcine that is stored at the Idaho Nuclear Technology and Engineering Center (INTEC) at Idaho National Laboratory. HIP trials performed with simulated calcines at Idaho National Laboratory’s Materials and Fuels Complex and an Australian Nuclear Science and Technology Organization facility from 2007 to 2010 produced a dense, monolithic waste form with increased chemical durability and effective (storage) volume reductions of ~10 to ~70% compared to granular calcine forms. In December 2009, the U.S. Department of Energy signed an amended Record of Decision selecting HIP technology as the treatment method for the 4,400 m3 of granular zirconia and alumina calcine stored at INTEC. Testing showed that HIP treatment reduces the risks associated with radioactive and hazardous constituent release, post-production handling, and long-term (repository) storage of calcines and would result in estimated storage cost savings in the billions of dollars. Battelle Energy Alliance has the ability to complete pilot-scale HIP processing of INTEC calcine, which is the next necessary step in implementing HIP processing as a calcine treatment method.

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

Performance Assessment for Closure of the Calcined Solids Storage Facility at the INL Site - 19289

The Calcined Solids Storage Facility is composed of six discrete stainless steel 'bin sets' at the Idaho Nuclear Technology and Engineering Center at the INL Site. Approximately 4,440 m{sup 3} of the dry, granular calcine waste is stored in the facility. Closure of the facility will entail removal of the calcine waste followed by grouting. A 5-cm thick residual amount of calcine was assumed to remain at the bottom of the bins. Important radionuclides in the residual layer for the groundwater pathway were identified as the fission and activation products Tc-99, Se-79, I-129, Cs-135, H-3, and actinides Pu-239, and Pu-240. The stainless-steel bins prevent near-term release of grouted residual calcine waste to the environment. Future releases were conservatively based on the corrosion rate of the weld seams on the stainless-steel bins, which site-specific studies measured mean tunnel corrosion rates of 0.0132 mm/yr with a coefficient of variation of 0.333. Based on tunnel corrosion rates, mean bin hydrologic containment times were estimated to be between 238 years and 1,191 years, depending on the thickness of the bin. Any breach in the weld seam was assumed to result in water flow and radionuclide leaching through the bin. The structural integrity of the bin was assumed to be controlled by the general stainless-steel corrosion rate, which was much slower (3.5 x 10{sup -5} mm/yr), resulting in structural integrity times between 24,000 years to over 100,000 years. A mixing-cell model was used to compute water flow and transport through the hydrologically failed bins and the vadose zone. Credit was taken for the geochemical properties of the grout used to fill the bins at closure, which significantly slows the releases from the bins for key radionuclides like Tc-99. A semi-analytical model for three-dimensional dissolved-phase transport in an aquifer of constant unidirectional flow, infinite length and width, and finite thickness was used to compute concentrations in the aquifer across the model domain for a time varying source. Cumulative impacts from all bin sets were calculated by summation across the bin set sources that were superimposed on a uniform Cartesian grid. Effective doses were calculated for direct groundwater ingestion from wells located 100-m downgradient from each of the bin sets and ingestion of meat, milk, and vegetables irrigated with water derived from the well. Intruder doses were based on a well drilled through the grouted waste and cuttings spread around a home site after a 500-year institutional control period. Peak doses during the 1,000-year compliance period were less than 5% of the 250 μSv/yr (25 mrem/yr) performance objective and intruder doses were generally less 5% of the of the 5000 μSv/yr (500 mrem) and 1000 μSv/yr (100 mrem) performance objective for acute and chronic intruder exposure, respectively. Groundwater concentrations were all less than the maximum contaminant limits. In all cases, doses were highest for the oldest bin set because, compared to the other bin sets, it had the highest radionuclide concentrations in the calcine waste. The performance assessment results indicate that performance objectives can be met leaving considerably more than 5-cm of residual calcine waste in the bin sets. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1961

Laboratory Development of a Pot Calcination Process for Converting Liquid Wastes to Solids

Simulated Darex, Purex, Thorex, and TBP-25 wastes were batch-evaporated and calcined to solids. The weight of residue was decreased to ~ 8% of the original waste for Darex, to ~ 4% for Purex, to ~ 2% for Thorex, and ~ 7% for TBP-25 by calcination to 1200°C. Semicontinuous evaporation and calcination to solids at 900°C of Darex and Purex wastes gave volume reduction factors of ~ 8 and 7 to 10, respectively. The nitrate contents of the residues from batch calcination to 800°C were from ~ 0.07 to 0.5 wt % while the nitrate contents of the residues from semicontinuous calcination to 900°C were from ~ 0.1 to 0.7 wt %. Sodium, calcium, and magnesium additives to Purex waste decreased the percent of sulfate volatilized into the condensate from ~ 30% to <0.5%. Replacement of the atmosphere above the waste with nitric oxide decreased ruthenium volatility from 50 to 60% of that originally present in Purex waste to 0.5 to 3.5%. The thermal conductivities for the calcined wastes, measured in situ, were all >0.1 Btu/hr ft °F at ~ 400°F and increased almost linearly with increasing temperature to >0.3 Btu/hr ft °F at ~ 1600°F in all cases studied.

052001* -- Nuclear Fuels-- Waste Processing↗
DOE OSTI2018

Calcined clay limestone cements (LC3)

The use of supplementary cementitious materials (SCMs) to replace part of the clinker in cement is the most successful strategy to reduce CO{sub 2} emissions in the global cement industry. However, limited supplies of conventional SCMs make it difficult to take this strategy further unless new types of SCMs become available. The only type of material available in the quantities needed to meet demand is clay containing kaolinite, which can be calcined to produce an effective SCM. Such clays are widely available in countries where most growth in demand for cement is forecast. Calcined clays have previously been used as pozzolans, but calcination makes the economics of substitution marginal in a conventional pozzolanic blend. The major innovation presented here is the possibility to make a coupled substitution of cement with calcined clay and limestone. This allows much higher levels of substitution. Blends where calcined clay is used as a pozzolan, typically have clinker contents around 65–70%. Combination of calcined clay with limestone allows higher levels of substitution down to clinker contents of around 50% with similar mechanical properties and improvement in some aspects of durability. The replacement of clinker with limestone in these blends lowers both the cost and the environmental impact.

36 MATERIALS SCIENCE↗