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At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7 waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.

36 MATERIALS SCIENCE↗
DOE OSTI2022

ETF Grout Expansion: Effects of Different Slags and Struvite Precipitation Protocols (Stirring Time)

The ETF ammonia tolerant grout waste form was developed by the Vitreous State Laboratory (VSL) to stabilize ammonium and thereby prevent emission of ammonia vapor during solidification of an ammonium-rich, concentrated sodium sulfate aqueous waste stream generated at the Hanford Effluent Treatment Facility (ETF). VSL personnel did not observe expansion in samples prepared during the waste form development work. However, Savannah River National Laboratory (SRNL) personnel detected expansion, in a few ETF grout samples, up to ~ 20 percent vertical expansion, while preforming work scope to evaluate the effect of ETF brine compositional ranges on the precipitation and solidification processes. The initial work requests can be found in Washington River Protection Services (WRPS) Statement of Work (SOW, Requisitions #: 339922 Revisions 0 and 1, March 11, 2021, and November 2021, respectively. As a result of the waste form expansion observed by SRNL, additional evaluation was requested by WRPS SOW Requisitions #: 356730 Revision 0 and Revision 1, February 1, 2022, and July 11, 2022, respectively. The goal of these requests was to determine whether differences in the slags and / or in sample preparation protocols used by VSL and SRNL were the cause of the expansion observed in the SRNL samples. G. Chen, WRPS, arranged a materials exchange between VSL and SRNL and coordinated the VSL protocol(s) for grouting and curing 1L batches of Base Case ETF base case simulant. Materials were exchanged between VSL and SRNL. Both laboratories performed the VSL struvite precipitation-solidification protocol and cured samples for at least 28 days. Differences in VSL and SRNL slag compositions, mineralogy and particle size results in themselves or combined were determined to not be the cause of the expansion because all samples expanded regardless of whether the VSL or SRNL slags were used. The fundamental cause of the observed expansion could not be attributed to minor variations in the following: 1) chemistry of the slag and other reagents, 2) container wall thickness, 3) ambient bench top curing conditions at VSL and SRNL, 4) pH adjustments to 7 ± a few tenths of unit, nor 5) stirring time between the precipitation and solidification process steps. Additional observations include: grout samples prepared with VSL slag expanded less than those prepared with the SRNL slag and, based on a very limited sample set, some samples stirred for longer times expanded less than those stirred for 20 minutes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Elimination of the Characterization of DWPF Pour Stream Sample and the Glass Fabrication and Testing of the DWPF Sludge Batch Qualification Sample

A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In contrast, the variability study has significantly added value to the DWPF's qualification strategy. The variability study has evolved to become the primary aspect of the DWPF's compliance strategy as it has been shown to be versatile and capable of adapting to the DWPF's various and diverse waste streams and blending strategies. The variability study, which aims to ensure durability requirements and the PCT and chemical composition correlations are valid for the compositional region to be processed at the DWPF, must continue to be performed. Due to the importance of the variability study and its place in the DWPF's qualification strategy, it will also be discussed in this report. An analysis of historical data and Production Records indicated that the recommendation of the Six Sigma team to eliminate all characterization of pour stream glass samples and the glass fabrication and PCT performed with the qualification glass does not compromise the DWPF's current compliance plan. Furthermore, the DWPF should continue to produce an acceptable waste form following the remaining elements of the Glass Product Control Program; regardless of a sludge-only or coupled operations strategy. If the DWPF does decide to eliminate the characterization of pour stream samples, pour stream samples should continue to be collected for archival reasons, which would allow testing to be performed should any issues arise or new repository test methods be developed.

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

Support of SRS salt-stone regulatory compliance - 15586

The SRS salt-stone Facility is designed and permitted to immobilize and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site. The salt-stone Processing Facility (SPF) is permitted as a wastewater treatment facility per the South Carolina Department of Health and Environmental Control Regulations (SCDHEC) South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-67. The salt-stone Disposal Facility is permitted as a solid waste landfill site, as defined by SCDHEC Regulations R61-107.19. The salt-stone Sampling and Analysis Plan provides the SCDHEC with the chemical and physical characterization strategy for the salt solution and salt-stone waste form. During operation, samples are collected from Tank 50H on a quarterly basis and are analyzed by the Savannah River National Laboratory (SRNL) for over chemical and radioactive components. The analytical results are compared to the salt-stone waste acceptance criteria in a technical report. SRNL also prepares salt-stone samples from the quarterly sample and premix binder materials provided by the SPF to determine the non-hazardous nature of the grout to meet the requirements of the SCDHEC regulation for the 'Characteristic of Toxicity', R.61-79.261.24(b) and the 'Universal Treatment Standards', R.61-79.268.48.4. salt-stone samples are collected, size reduced, packaged, and shipped to an independent United States EPA certified laboratory, or a laboratory that holds an equivalent certification such as one from the National Environmental Laboratory Accreditation Program to perform the EPA Toxicity Characteristic Leaching Procedure (TCLP) extraction of the salt-stone sample. SRNL reviews the results of the TCLP analysis and prepares a report detailing the sample preparation and analysis that is used by the salt-stone Facility to demonstrate the nonhazardous nature of the salt-stone waste form. (authors)

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

Saltstone 1QCY15 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the first quarter of calendar year 2015 (1QCY15). After a 28 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP). The 1QCY15 saltstone sample results meet South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents (UHC).

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

Saltstone 2QCY15 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the second quarter of calendar year 2015 (2QCY15). After a 28 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP). The 2QCY15 saltstone sample results meet South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents (UHC).

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

Saltstone 3QCY15 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the third quarter of calendar year 2015 (3QCY15). After a 28 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP).1 The 3QCY15 saltstone sample results meet South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents (UHC).

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

Saltstone 2QCY15 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the second quarter of calendar year 2015 (2QCY15). After a 28 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP) 1 . The 2QCY15 saltstone sample results meet South Carolina Hazardous Waste Management Regulations (SCHWMR) R.61-79.261.24 and R.61-79.268.48(a) requirements for a nonhazardous waste form with respect to RCRA metals and underlying hazardous constituents (UHC).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Pelletization with Spark Plasma Sintering and Characterization of Metal Iodides: An Assessment of Long-Term Radioiodine Immobilization Options

Four promising iodine “getter” materials (Ag, Cu, Bi, and Sn) for radioiodine capture were assessed in their pure metal-iodide (MI x ) pelletized forms to compare relative chemical durabilities. To study chemical durability, commercial MI x compounds of AgI, BiI 3 , BiOI, CuI, and SnI 4 were converted to dense monolithic pellets using spark plasma sintering. Semidynamic leach testing in the form of modified ASTM C1308 tests was then performed on the pellets in two different forms including unmounted (as-pressed) specimens (i.e., “U”) and epoxy-mounted specimens (i.e., “M”) with polished surfaces. The chemical durability results and sample characterizations showed that three of the five MI x compounds tested (i.e., AgI, CuI, and BiOI) displayed moderate to high leach resistances. Further, the remaining two MI x compounds (i.e., BiI 3 and SnI 4 ), which are both desirable iodine waste forms due to their high iodine loading capacities, readily decomposed during leach testing, indicated by crystallographic changes in the specimens as well as large amounts of iodine detected in the leachate solutions. The instabilities of BiI 3 and SnI 4 raise uncertainties for using the base metals/cations (i.e., Bi 0 /Bi 3+ and Sn 0 /Sn 4+ , respectively) as viable getters for radioiodine capture due to likely poor waste form chemical durabilities after capture and consolidation into waste forms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Scientific Solutions to Nuclear Waste Environmental Challenges

The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount of time it takes for one-half of the material to undergo radioactive decay.) In general, the ideal material would need to be durable for approximately 10 half-lives to allow the activity to decay to negligible levels. However, the potential health effects of each radionuclide vary depending on what type of radiation is emitted, the energy of that emission, and the susceptibility for the human body to accumulate and concentrate that particular element. Consequently, actual standards tend to be based on limiting the dose (energy deposited per unit mass) that is introduced into the environment. The Environmental Protection Agency (EPA) has the responsibility to establish standards for nuclear waste disposal to protect the health and safety of the public. For example, the Energy Policy Act of 1992 directed the EPA to establish radiation protection standards for the Yucca Mountain geologic repository for nuclear wastes. The standards for Yucca Mountain were promulgated in 2008, and limit the dose to 15 millirem per year for the first 10,000 years, and 100 milirem per year between 10,000 years and 1 million years (40 CFR Part 197; http://www.epa.gov/radiation/yucca/2008factsheet.html). So, the challenge is two-fold: (1) develop a material (a waste form) that is capable of immobilizing the waste over geologic time scales, and (2) develop a process to convert the radioactive sludge in the tanks into this durable waste form material. Glass: Hard, durable, inert, and with infinite chemical versatility Molten glass is a powerful solvent liquid, which can be designed to dissolve almost anything. When solidified, it can be one of the most chemically inert substances known to man. Nature's most famous analogue to glass is obsidian, a vitreous product of volcanic activity; formations over 17 million years old have been found. Archaeologists have found man-made glass specimens that are five thousand years old.

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

Update on SRNL Support of Saltstone Sampling and Analysis - 16325

At the Savannah River Site, low-level waste is immobilized in cementitious waste form known as saltstone. The saltstone Performance Assessment models the performance of the waste form and other aspects of the Saltstone Disposal Facility over ten thousand years after closure to determine the transport history of radionuclides and other hazardous constituents in low-level waste. As part of an ongoing Performance Assessment Maintenance Plan, Savannah River Remediation with Savannah River National Laboratory, has designed a sampling and analyses plan to correlate the properties of field-emplaced samples and samples processed and cured in the laboratory. As part of the plan, SRR has retrieved field-emplaced samples from SDU 2A. Prior to analyzing the field emplaced samples, SRNL is developing the strategy, protocols and equipment necessary to retrieve, store in an inert environment, and analyze the three sample sets of interest outlined in the analysis plan. To date, the simulant saltstone samples have been analyzed for hydraulic conductivity, porosity, density, and other properties. This paper will give an update to the current protocols and equipment being utilized at Savannah River National Laboratory for storing and analyzing the field-emplaced and simulant samples, results from the simulant saltstone analyses, as well as analysis plans for the samples retrieved from the Saltstone Processing Facility and the field-emplaced samples. (authors)

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

Analysis of Saltstone Cores Taken from Saltstone Disposal Unit Cell 2A - 17350

At the Savannah River Site, low-level waste is immobilized in cementitious waste form known as saltstone. The saltstone Performance Assessment models the performance of the waste form and other aspects of the Saltstone Disposal Facility over ten thousand years after closure to better understand the transport history of radionuclides and other hazardous constituents in low-level waste. As part of ongoing Performance Assessment Maintenance, Savannah River Remediation (SRR), with Savannah River National Laboratory (SRNL), designed a sampling and analyses plan to correlate the properties of field-emplaced samples and samples processed and cured in the laboratory. The primary objectives of the Sampling and Analyses Plan are; (1) to demonstrate a correlation between the measured properties of laboratory-prepared, simulant samples (termed Sample Set 3), and the field-emplaced saltstone samples (termed Sample Set 9), and (2) to validate property values assumed for the Saltstone Disposal Facility (SDF) Performance Assessment modeling. The analysis and property data for Sample Set 9 (i.e. core samples extracted from Saltstone Disposal Unit Cell 2A (SDU2A)) are documented, and where applicable, the results are compared to the results for Sample Set 3. SRNL received, stored and analyzed the samples cored from SDU2A. This paper discusses the results of the SRNL analysis of the SDU2A samples as well as the simulant saltstone samples. (authors)

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

Solidification/Stabilization of Liquid Oil Waste in Meta-kaolin-based geo-polymer - 17209

Cement-base solidification/stabilization (S/S) is a widely applied technique for the immobilization of inorganic hazardous constituents. However, S/S is considered less compatible with organic liquid wastes (OLW) because organic compounds may inhibit binder hydration. geo-polymers (alumino-silicate materials) are a worthwhile material for applications in which cements are inappropriate. In this study, we propose a direct immobilization process of organic liquid waste in a geo-polymer matrix based on the emulsification of OLW in an activating solution, followed by the addition of aluminosilicate source (meta-kaolin). At first, the harmlessness of the chemicals contained in the OLW on the geo-polymerization was examined through a screening, reveling that only acidic oils should be pre-treated before a direct immobilization. Waste form characterizations with a simulated OLW were measured, showing interesting mechanical and confinement properties. Waste forms containing 20% Vol. of OLW show a mechanical strength higher than 30 MPa in compression and less than 1 % of waste leached over a month in pure water. (authors)

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

Novel inorganic sorbent for iodine removal in pump and treat system - 15020

The radioisotope {sup 129}I is one of the key risk components in the migrating subsurface plumes present in the Hanford Site vadose zone due to its long half-life, large inventory, high toxicity, and high mobility in the aquifer. Removal of iodine from groundwater in the 200 Area is an essential component for managing the Hanford Site cleanup and protecting the Columbia River. In the subsurface aquatic environment it mainly exist as iodide, iodate, and organic derivatives of iodine. Because of different chemical properties of these species and significantly lower concentration than other constituents in the subsurface plumes their removal using commercial ion exchange resins is problematic. In addition, availability of a stable matrix for a long-term waste form for iodine is needed. Novel engineered inorganic composites that possess superior characteristics including large surface area, high sorption capacity, and good chemical-mechanical-thermal stability may potentially offer a solution. These two-in-one novel materials can be designed to serve as an efficient sorbent for the simultaneous removal of the dissimilar iodine species and a durable waste form under environmentally relevant conditions. This paper describes our current results on identification, characterization, and testing of the iodine-selective inorganic sorbents. (authors)

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

Radiation Stability of Spark‐Plasma‐Sintered Lead Vanadate Iodoapatite

Spark‐plasma‐sintered lead vanadate iodoapatite Pb 9.85 (VO 4 ) 6 I 1.7 , a promising nuclear waste form for the immobilization of I‐129, was irradiated with energetic ions, electrons, and gamma rays, to investigate its radiation stability. In situ TEM observation of the 1 MeV Kr 2+ irradiation shows that lead vanadate iodoapatite generally exhibits higher tolerance against ion irradiation‐induced amorphization than lead vanadate fluorapatite, and the spark plasma sintering can further enhance its radiation stability attributed to the enhanced crystallinity, reduced defect concentration, and denser microstructure. The critical amorphization dose and critical temperature for the SPS‐densified iodoapatite at 700°C are determined to be 0.25 dpa at room temperature and 230°C, respectively. No significant phase transformation or microstructural damage occurred under energetic electron and gamma irradiations. Raman spectra of gamma‐ray‐irradiated iodoapatite indicate improved V–O bond order at 500 kGy dose. Generally, the spark‐plasma‐sintered iodoapatite exhibits excellent radiation stability for nuclear waste form applications. The significantly enhanced radiation stability of the SPS‐densified iodoapatite suggests that SPS holds great promise for fabricating iodoapatite waste form with minimum iodine loss and optimized radiation tolerance for effective management of highly volatile I‐129.

Lu, Fengyuan↗
DOE OSTI2011

Technical Insights for Saltstone PA Maintenance

The Cementitious Barriers Partnership (CBP) is a collaborative program sponsored by the US DOE Office of Waste Processing. The objective of the CBP is to develop a set of computational tools to improve understanding and prediction of the long-term structural, hydraulic, and chemical performance of cementitious barriers and waste forms used in nuclear applications. CBP tools are expected to better characterize and reduce the uncertainties of current methodologies for assessing cementitious barrier performance and increase the consistency and transparency of the assessment process, as the five-year program progresses. In September 2009, entering its second year of funded effort, the CBP sought opportunities to provide near-term tangible support to DOE Performance Assessments (PAs). The Savannah River Saltstone Disposal Facility (SDF) was selected for the initial PA support effort because (1) cementitious waste forms and barriers play a prominent role in the performance of the facility, (2) certain important long-term behaviors of cementitious materials composing the facility are uncertain, (3) review of the SDF PA by external stakeholders is ongoing, and (4) the DOE contractor responsible for the SDF PA is open to receiving technical assistance from the CBP. A review of the current (SRR Closure & Waste Disposal Authority 2009) and prior Saltstone PAs (e.g., Cook et al. 2005) suggested five potential opportunities for improving predictions. The candidate topics considered were (1) concrete degradation from external sulfate attack, (2) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, (3) mechanistic prediction of geochemical conditions, (4) concrete degradation from rebar corrosion due to carbonation, and (5) early age cracking from drying and/or thermal shrinkage. The candidate topics were down-selected considering the feasibility of addressing each issue within approximately six months, and compatibility with existing CBP expertise and already-planned activities. Based on these criteria, the five original topics were down-selected to two: external sulfate attack and mechanistic geochemical prediction. For each of the selected topics, the CBP communicated with the PA analysts and subject matter experts at Savannah River to acquire input data specific to the Saltstone facility and related laboratory experiments. Simulations and analyses were performed for both topics using STADIUM (SIMCO 2008), LeachXS/ORCHESTRA (ECN 2007, Meeussen 2003), and other software tools. These supplemental CBP analyses produced valuable technical insights that can be used to strengthen the Saltstone PA using the ongoing PA maintenance process. This report in part summarizes key information gleaned from more comprehensive documents prepared by Sarkar et al. (2010), Samson (2010), and Sarkar (2010).

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

Atomistic scale investigation of cation ordering and phase stability in Cs-substituted Ba 1.33 Zn 1.33 Ti 6.67 O 16 , Ba 1.33 Ga 2.66 Ti 5.67 O 16 and Ba 1.33 Al 2.66 Ti 5.33 O 16 hollandite

The titanate-based hollandite structure is proposed as an effective ceramic waste form for Cs-immobilization. In this study, quantum-mechanical calculations were used to quantify the impact of A-site and B-site ordering on the structural stability of hollandite with compositions Ba x Cs y (M z Ti 8-z )O16, where M = Zn 2+ , Ga 3+ , and Al 3+ . The calculated enthalpy of formation agrees with experimental measurements of related hollandite phases from melt solution calorimetry. Ground state geometry optimizations show that, for intermediate compositions (e.g., CsBaGa 6 Ti 18 O 48 ), the presence of both Cs and Ba in the A-site tunnels is not energetically favored. However, the decay heat generated during storage of the Cs-containing waste form may overcome the energetics of Ba and Cs mixing in the tunnel structure of hollandite. The ability of the hollandite structure to accommodate the radioparagenesis of Cs to Ba is critical for long term performance of the waste. For the first time, B-site ordering was observed along the tunnel direction ([001] zone axis) for the Ga-hollandite compositions, as well as the intermediate Al-hollandite composition. These compositionally dependent structural features, and associated formation enthalpies, are of importance to the stability and radiation damage tolerance of ceramic waste forms.

36 MATERIALS SCIENCE↗