Atomic and micro-structure features of nanoporous aluminosilicate glasses from reactive molecular dynamics simulations
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Materials Science↗
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Spectroscopic identification of Ca-bearing uranyl silicates formed in C–S–H systems
Portland cement-based grouts used for radioactive waste immobilisation contain a Ca- and Si-rich binder phase, known as calcium–silicate–hydrate (C–S–H). Depending on the blend of cement used, the Ca/Si ratio can vary considerably. A range of C–S–H minerals with Ca/Si ratios from 0.6 to 1.6 were synthesised and contacted with aqueous U(VI) at 0.5 mM and 10 mM concentrations. Solid-state 29 Si MAS-NMR spectroscopy was applied to probe the Si coordination environment in U(VI)-contacted C–S–H minerals and, in conjunction with U L III -edge X-ray absorption spectroscopy analysis, inferences of the fate of U(VI) in these systems were made. At moderate or high Ca/Si ratios, uranophane-type uranyl silicates or Ca-uranates dominated, while at the lowest Ca/Si ratios, the formation of a Ca-bearing uranyl silicate mineral, similar to haiweeite (Ca[(UO 2 ) 2 Si 5 O 12 (OH) 2 ]·3H 2 O) or Ca-bearing weeksite (Ca 2 (UO 2 ) 2 Si 6 O 15 ·10H 2 O) was identified. This study highlights the influence of Ca/Si ratio on uranyl sequestration, of interest in the development of post-closure safety models for U-bearing radioactive waste disposal.
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Solubility controls on plutonium and americium release in subsurface environments exposed to acidic processing wastes
To identify the role of waste composition and sediment interactions in controlling Pu and Am mobility in contaminated sediments at Hanford, a legacy nuclear site, Pu and Am concentrations in solutions equilibrated with contaminated sediments from beneath the 216-Z-9 (Z-9) Trench were compared to the solubilities of PuO 2 materials, synthesized by methods representative of the disposed wastes, in the absence of sediments. Furthermore, this work shows that the solubilities of PuO 2 materials synthesized by different methods, and with varying particle sizes, agree with PuO 2 (am,hyd), although dissolution kinetics differed between materials. According to saturation index (SI) calculations, PuO 2 (am,hyd) is likely also controlling Pu release from sediments under conditions where phosphate concentrations are low. However, both Pu-phosphate and Am-phosphate phases, identified in SI calculations and by high resolution transmission electron microscopy, play roles in controlling release in low pH, high phosphate, shallow sediments just below the Z-9 Trench. The elevated phosphate is likely due to decomposition of tributyl phosphate from waste solutions over time. Sediments from deeper in the subsurface beneath the Z-9 Trench are less acidic and contain less phosphate, with Pu solubility likely controlled by PuO 2 (am,hyd) that precipitated following neutralization of the acidic waste stream. Controls on Am concentrations in deeper sediments are more complex and potentially involve sediment adsorption and/or release from Pu (1-x) Am x O 2 following Am in-growth. The concentrations of both Pu and Am were elevated in the colloidal fraction associated with shallow sediments, but not in PuO 2 experiments, suggesting the presence of Pu/Am pseudocolloids (e.g., Pu/Am associated with mineral colloids). However, Pu and Am association with the colloidal size fraction was not observed in deeper sediments, suggesting transport of Pu and Am to these depths beneath the Z-9 Trench was not due to colloidal transport.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Adaptation of the NWM Cloud Environment for an ISF Project
The DOE-NE NWM Cloud was designed to be a generic set of tools and applications for any nuclear waste management program. As policymakers continue to consider approaches that emphasize consolidated interim storage and transportation of spent nuclear fuel, a gap analysis of the tools and applications provided for spent nuclear fuel and high-level radioactive waste disposal in comparison those needed for siting, licensing, and developing a consolidated interim storage facility and/or for a transportation campaign will help prepare DOE for implementing such potential policy direction. This report evaluates the points of alignment and potential gaps between the applications on the NWM Cloud that supported SNF disposal project, and the applications needed to address QA requirements and for other project support needs of an SNF storage project.
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Method Development and Implementation for Characterizing Uncertainties in Tank Leak Volumes for Composite Analysis.
The purpose of this environmental calculation file (ECF) is to provide a summary of methodology, inputs, assumptions, and results of a Special Analysis (SA) titled “Method Development and Implementation for Characterizing Uncertainties in Tank Leak Volumes for the Updated Composite Analysis.” This SA is one of the three SAs conducted in Fiscal Year (FY) 2024 for the Composite Analysis (CA) for the Hanford Site (DOE/RL-2019-52, Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2020)), and it is performed following the Unreviewed Composite Analysis Question Evaluation procedure, UCAQ-24-01-E (UCAQ-24-01-E, Rev. 0).
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Effects of Mineral Impurities and Heat on Uranium(VI) Sorption onto Bentonite (Final Report)
Most nuclear waste disposal options include compacted bentonite, consisting of montmorillonite clay, as part of a barrier system to minimize contaminant mobility. Uranium (U) is the primary element in spent nuclear fuel, and from an environmental perspective, a potential contaminant of water resources. Furthermore, uranium may control nuclear fuel degradation rates and the consequent release of other radioactive contaminants based on its diffusive transport away from waste canisters. Uranium sorption onto clay and other mineral surfaces is expected to limit U(VI) mobility in these systems. However, at this point a prediction of U(VI) sorption and transport behavior in performance assessment (PA) models is complicated by a series of factors, such as: (1) the presence of various U(VI) solution species with different charges and sorption characteristics; (2) the complex microstructure of montmorillonite clay resulting in two types of clay porosities and multiple sorption sites; (3) the largely unknown effects of bentonite mineral impurities on pore water chemistry, U(VI) sorption and diffusion behavior; and (4) the potential impacts of heat, generated by the decay of spent fuel, on mineralogical and microstructural transformations, and any subsequent effects on radionuclide sorption and mobility. In this project, we investigated the effects of calcite impurities on U(VI) sorption and diffusion onto montmorillonite before and after mineral exposure to heat. Overall, our main goals were to: 1. Gain a mechanistic understanding of the effects of calcite impurities on U(VI) sorption and diffusion in montmorillonite/bentonite at variable chemical conditions, and before and after mineral exposure to heat; and to 2. Work towards the development of new surface complexation and diffusion models to predict U(VI) sorption and mobility in montmorillonite/bentonite while accounting for the impacts of calcite impurities and heat. In order to achieve these goals, we pursued the following research objectives and tasks: 1. Characterize the effects of calcite impurities on U(VI) sorption onto Na-montmorillonite, and evaluate if and how these effects change due to solids’ exposure to heat (Task 1). 2. Characterize U(VI) surface speciation in solid samples from Task 1, with a particular focus on U(VI) precipitates and co-precipitates (Task 2). 3. Test the relevance of observed changes in U(VI) sorption characteristics for U(VI) diffusive transport behavior in bentonite barriers (Task 3). 4. Evaluate the exclusion of anionic U(VI) solution complexes from clay nanopores (Task 4). 5. Building on Tasks 1-4, start the development of new surface complexation and reactive transport models that can simulate and predict U(VI) sorption and diffusion in complex bentonite systems (Task 5). A detailed description of all tasks, and their specific results and conclusions is provided in this report.
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Annual Status Report (FY 2020): Performance Assessment for the Disposal of Low Level Waste in the 200 East Area Burial Grounds
This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730). The estimates are compared with requirements of DOE O 435.1 Chg 1 and performance objectives defined in companion documents DOE M 435.1-1 Chg 1and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. A new PA study was initiated in fiscal year (FY) 2019 for evaluation of active disposal sites within the 200 East and 200 West Areas (Trench 94 in 200 East; Trenches 31 and 34 in 200 West) due to extended time elapsing between the current annual status report and the original PA for the active disposal sites. The new PA for the active disposal sites is expected to be completed in FY 2021. Within the active burial grounds in the 200 East Area, low-level waste and mixed low-level waste will continue to be disposed of in the dedicated U.S. Navy reactor compartment trench at the 218-E-12B Burial Ground (Trench 94). During this reporting period (FY 2020, from October 1, 2019, through September 30, 2020), two reactor compartments were disposed in Trench 94. Results from sorption experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., technetium-99 and iodine-129). The test durations ranged from 1 to 3 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East and 200 West Areas in the Central Plateau (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-707) estimates that the 200 East LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1 Chg 1. Overall, there are no substantive changes to primary PA assumptions nor the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 and the Disposal Authorization Statement is maintained.
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Microbial interactions with phosphorus containing glasses representative of vitrified radioactive waste
The presence of phosphorous in borosilicate glass (at 0.1 – 1.3 mol % P 2 O 5 ) and in iron-phosphate glass (at 53 mol % P 2 O 5 ) stimulated the growth and metabolic activity of anaerobic bacteria in model systems. Dissolution of these phosphorous containing glasses was either inhibited or accelerated by microbial metabolic activity, depending on the solution chemistry and the glass composition. The breakdown of organic carbon to volatile fatty acids increased glass dissolution. The interaction of microbially reduced Fe(II) with phosphorus-containing glass under anoxic conditions decreased dissolution rates, whereas the interaction of Fe(III) with phosphorous-containing glass under oxic conditions increased glass dissolution. Phosphorous addition to borosilicate glasses did not significantly affect the microbial species present, however, the diversity of the microbial community was enhanced on the surface of the iron phosphate glass. Results demonstrate the potential for microbes to influence the geochemistry of radioactive waste disposal environments with implication for wasteform durability.
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CA Special Studies: Updates to the Groundwater Pathway Radiological Dose
This environmental calculation file (ECF) summarizes the groundwater pathway exposure assumptions and methodologies used to revise radionuclide specific unit dose factors (UDFs) published in CP-64491, Hanford Site Composite Analysis Data Package – Exposure Scenarios and Radionuclide Specific Dose Conversion Factors. Specifically, this ECF describes the updated exposure-related parameters used to calculate radionuclide specific UDFs based on recommendations from the Low-Level Waste Disposal Facility Federal Review Group (LFRG) and updates to Adult External DCFs included in EPA 402-R-93-081, External Exposure to Radionuclides in Air, Water, and Soil hereinafter called Federal Guidance Report [FGR]).
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Proceedings of the 13 th US/German Workshop on Salt Repository Research, Design, and Operation
This report summarizes the proceedings of the 13 th US/German Workshop on Salt Repository Research, Design, and Operation hosted by Sandia National Laboratories on June 20-23, 2023, in Santa Fe, New Mexico, USA. Over 60 participants attended, representing Germany, United States, the Netherlands, Australia, and the United Kingdom, along with the IAEA. The purpose of the US/German Workshop is to foster in-person collaboration and dialogue amongst salt repository researchers and nuclear waste disposal implementers across international organizations. The workshop included five sessions of topical presentations and two breakout sessions to promote additional discussion on compelling topics.
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Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028
Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)
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National and regional waste stream in the United States: conformance and disparity
Accurate estimation of municipal solid waste (MSW) composition is critical for efficient waste management. In the United States, site-specific and material flow approaches determine the MSW composition at regional and national levels. The material flow-based national estimates are determined by the US EPA; the US EPA’s estimates are known to differ substantially from the aggregated tonnage of MSW managed by waste handling facilities in the United States. However, the material class-specific discrepancies of the US EPA’s material flow approach resulting in these differences are unknown. To find the basis of these discrepancies, we analyze the discarded MSW stream of 27 US states, which roughly accounts for 73 percent of the US population. Our analysis indicates that the material flow-based national estimates are accurate for the food, plastic, and glass material classes. In contrast, we find that the US EPA’s material flow-based predictions underestimate paper waste disposal by at least 15 million tons annually. These differences likely stem from incorrect assumptions of residence time. These results highlight the material class-specific strengths and drawbacks of the US EPA’s material flow-based MSW estimates.
29 ENERGY PLANNING, POLICY, AND ECONOMY↗
Polyethylene upcycling to long-chain alkylaromatics by tandem hydrogenolysis/aromatization
The current scale of plastics production and the accompanying waste disposal problems represent a largely untapped opportunity for chemical upcycling. Tandem catalytic conversion by platinum supported on γ-alumina converts various polyethylene grades in high yields (up to 80 weight percent) to low-molecular-weight liquid/wax products, in the absence of added solvent or molecular hydrogen, with little production of light gases. The major components are valuable long-chain alkylaromatics and alkylnaphthenes (average ~C 30 , dispersity Ð = 1.1). Coupling exothermic hydrogenolysis with endothermic aromatization renders the overall transformation thermodynamically accessible despite the moderate reaction temperature of 280°C. This approach demonstrates how waste polyolefins can be a viable feedstock for the generation of molecular hydrocarbon products.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Sorption of Metals and Radionuclides in Cementitious Leachate Impacted Sediments from the Savannah River Site
Performance assessments (PAs) are calculations used to establish the risk posed by radioactive waste disposal facilities in the Savannah River Site (SRS). The distribution coefficient (Kd; solid/liquid concentration ratio; Csolid/liquid) is one of the key geochemical parameters used in the PA to provide a measure of how strongly the dissolved contaminants bind to solid phases; the greater the Kd value, the greater the tendency to bind to the solids. The overall purpose of this study was to measure this key parameter for several constituents of concern in the cementitious leachate impacted SRS sediment environments. Particular attention was directed at quantifying the influence of cementitious leachate from various stages of cement aging on the sorption of radioactive waste and heavy metals on SRS sediments.
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Evaluation of Engineered Barrier Systems (FY2022 Report)
This report describes research and development (R&D) activities conducted during Fiscal Year 2022 (FY22) specifically related to the Engineered Barrier System (EBS) R&D Work Package in the Spent Fuel Waste Science and Technology (SFWST) Campaign supported by the United States (U.S.) Department of Energy (DOE). The R&D activities focus on understanding EBS component evolution and interactions within the EBS, as well as interactions between the host media and the EBS. The R&D team represented in this report consists of individuals from Sandia National Laboratories, Lawrence Berkeley National Laboratory (LBNL), Los Alamos National Laboratory (LANL), and Vanderbilt University. EBS R&D work also leverages international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear Safety [Vol. 16, No. 2, March-April 1975]
Nuclear Safety covers significant developments in the field of nuclear safety. The scope is limited to topics relevant to the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, safety considerations in regard to the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 127 Quality Assurance in the Construction of Nuclear Power Plants by Sidney A. Bernsen, 141 1974 ANS Topical Meeting on Fast Reactor Safety by M. H. Fontana; CONTROL AND INSTRUMENTATION: 150 GBR-4 Protection Systems: Failures and Their Consequences by Peter Burgsmüller, J. J. Dekais, Albert Krähe, Raffaello Pignatelli, and Gottfried Vieider, 162 Standby Emergency Power Systems. Part 2—The Later Plants by E. W. Hagen; PLANT SAFETY FEATURES: 180 Radiotoxic Hazard Measure for Buried Solid Radioactive Waste by J. Hamstra, 190 The Thirteenth AEC Air-Cleaning Conference by D. W. Moeller, D. W. Underhill, and M. W. First, 203 Book Review: Nuclear Criticality Safety; CONSEQUENCES OF EFFLUENT RELEASE: 204 Environmental Radiation Effects of Nuclear Facilities in New York State by M. S. Terpilak and B. L. Jorgensen, 222 Book Review: Thermal Ecology; OPERATING EXPERIENCES: 223 Set-Point Drift in Nuclear Power-Plant Safety-Related Instrumentation Adapted by the Nuclear Safety Staff, 224 Diesel-Generator Operating Experience at Nuclear Power Plants, 227 Summary of Operating U. S. Power Reactors as of Jan. 1, 1975, 232 Selected Safety-Related Occurrences Reported in November and December 1974 Compiled by William R. Casto, 235 Recent Occurrences at Nuclear Reactors and Their Causes Compiled by William R. Casto; CURRENT EVENTS: 243 General Administrative Activities Compiled by Wm. B. Cottrell, 251 Action on Power-Reactor Projects Undergoing Regulatory Review or Consideration Compiled by Wm. B. Cottrell, 266 Action on Nonreactor Projects Undergoing Regulatory Review or Consideration Compiled by Wm. B Cottrell, 268 Proposed Rule Changes as of Jan. 1, 1975; MISCELLANY: 149 Course in Italy on High-Energy Radiation Dosimetry and Protection (Announcement), 250 Course at Northwestern on Safety of Light-Water-Cooled Nuclear Power Plants (Announcement), 266 Symposium of the Combined Effects on the Environment of Radioactive, Chemical, and Thermal Releases from the Nuclear Industry (Announcement), 271 Short Course on Engineering for Extreme Winds and Tornadoes (Announcement), 272 Three 1-Week Courses at MIT on Nuclear Power-Reactor Safety (Announcement), 272 Harvard University Short Courses (Announcement).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PFAS Removal by Ion Exchange Resins: Background and Knowledge Gaps with Respect to the Hanford Site
Per- and polyfluoroalkyl substances (PFAS) have been a rising concern for the past two decades, with the United States Department of Defense and Environmental Protection Agency investing millions of dollars in research into remediation and clean-up technologies. Due to the environmental persistence, toxicity, biological uptake, and ongoing changes in both federal and state regulatory space, understanding the fate and transport of PFAS compounds has been of growing concern to the US Department of Energy (DOE). The DOE’s Hanford Site is investigating historical use of PFAS and will be doing site characterization for PFAS. Thus, PFAS have not yet been identified as a contaminant concern in regulatory documents. Based on historical records that mention the discharge of aqueous film-forming foam containing PFAS and having on-site fire stations (a risk factor for PFAS contamination), it seems likely that environmental releases of PFAS may have occurred. Pump and treat (P&T) remediation is the selected remedy for multiple groundwater contaminant plumes at Hanford. These P&T systems use ion exchange (IX) as a component of aboveground treatment, with the specific resins depending on the target contaminants. There is potential that these IX resins may be able to remove PFAS from groundwater, but investigation is needed to understand affinity/selectivity and removal capacity given the groundwater composition and the operating conditions. This report provides background on PFAS uses and chemistry, then provides a review of IX resin applications for PFAS, identifying knowledge gaps. Recommendations are provided regarding research needed to address knowledge gaps and acquire information needed to propose IX as a future PFAS remediation technology at the Hanford Site, as well as other U.S. Department of Energy sites. Generally, PFAS compounds are fluorinated substances that contain at least one fully fluorinated methyl or methylene carbon – with a few noted exceptions, any chemical with at least a perfluorinated methyl group (–CF3) or a perfluorinated methylene group (–CF2–) is a PFAS. These chemical compounds are characterized as non-biodegradable, non-reactive, non-photolytic, and hydrolysis resistant. This makes them highly recalcitrant within the environment, however polyfluoroalkyl materials are less recalcitrant as the carbon chains contain C–H bonds which are more easily broken than carbon – fluorine (C–F) bonds. The backbone carbon structures are commonly punctuated with a head group, the most well-known of them are perfluorooctanesulfonic acid and perfluorooctanoic acid, which possess a sulfonate and a carboxylate group, respectively. IX resins are marketed for the removal of PFAS from water systems and industrial water, however, the mechanism of removal is not as well understood as for anion or cation removal. A better understanding of the mechanism of removal would enable the development of IX resins that have improved specificity for PFAS removal. Four knowledge gaps were identified: 1) the effect of dissolved ions on the IX resin PFAS removal effectiveness, 2) the effect of additional primary contaminants of concern (PCOCs) or secondary contaminants of concern (SCOCs) on the effectiveness of PFAS via IX resin, 3) the mechanisms of PFAS removal from water, and 4) practical solutions to IX resin regeneration and waste disposal.
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Updates to Composite Analysis Base Case and Null Space Monte Carlo Sensitivity Based on New Unit Dose Factors
This environmental calculation file (ECF) presents the results of the exposure route-specific and total radiological dose assessments for the groundwater exposure pathway as a part of the updated Composite Analysis for Low-Level Waste Disposal in the Hanford Site Central Plateau (FY 2022), based on revised unit dose factors (UDFs) published in CA Special Studies: Updates to the Groundwater Pathway Radiological Dose. The reason for these radiological dose assessments is to estimate the cumulative radiological impacts from all screened sources of ionizing radiation and exposure routes that could potentially contribute to the projected dose to a hypothetical member of the public. Sources of ionizing radiation can include existing or future disposal facilities and other sources including past-practice discharge sites. Additionally, the results of the exposure route-specific and total radiological dose assessments have been revised for the groundwater pathway based on the null space Monte Carlo (NSMC) groundwater concentrations and the most recent updates to the UDFs as a part an uncertainty analysis for the updated Hanford Site CA.
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