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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

'Radiotoxicity Index': An Inappropriate Discriminator for Advanced Fuel Cycle Technology Selection - 12276

A radiotoxicity index (RI) is often used as a figure of merit for evaluating for evaluating the attractiveness of employing an advanced fuel cycle (i.e., a fuel cycle that uses some combination of separations and other reactor technologies, such as fast reactors), rather than continued use of the current 'once-through' fuel cycle. The RI is calculated by multiplying the amount of every radionuclide found in a waste form for some unit amount of waste times the drinking water dose conversion factor, DCF, for each radionuclide, then summing these together. Some argue that if the RI for an advanced fuel cycle is lower than the RI for a once-through fuel cycle, then implementation of the particular advanced fuel cycle has merit because it reduces the radiotoxicity of the waste. Use of an RI for justifying separations technologies and other components of advanced fuel cycles is not only inappropriate, but can be misleading with respect to judging benefits of advance fuel cycle options. The disposal system, through its use of multiple engineered and natural barriers to migration, eliminates most of the radionuclides contributing to the RI such that additional separations technologies will make little difference to peak dose rates. What must also be considered is the health/dose risk caused to workers and the public by the construction and operation of the separations facility itself. Thus, use of RI may lead to selection of separations technologies that may have a negligible effect on lowering the potential health risks associated with disposal, but will increase real worker and public health risks in the near term. The use of the radiotoxicity index (RI) as a figure of merit for justifying advanced fuel cycles involving separations technologies is not only inappropriate, but can be misleading with respect to judging benefits of advance fuel cycle options. The disposal system, through its use of multiple engineered and natural barriers to migration, eliminates most of the radionuclides contributing to the RI such that additional separations technologies will make little difference to peak dose rates. What must also be considered is the health/dose risk caused to workers and the public by the construction and operation of the separations facility itself. Thus, use of RI may lead to selection of separations technologies that may have a negligible effect on lowering the potential health risks associated with disposal, but will increase real worker and public health risks in the near term. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Determination of Desorbed Species During Heating of AgI-Mordenite Provided by ORNL

This study is focused on describing the desorbed off gases due to heating of the AgIMordenite (MOR) produced at ORNL for iodine (I 2 ) gas capture from nuclear fuel aqueous reprocessing. In particular, the interest is for the incorporation of the AgI-MOR into a waste form, which might be the Sandia developed, low temperature sintering, Bi-Si oxide based, Glass Composite Material (GCM). The GCM has been developed as a waste form for the incorporation any oxide based getter material. In the case where iodine may be released during the sintering process of the GCM, additional Ag flake is added as further insurance in total iodine capture and retention. This has been the case for the incorporated ORNL developed AgIMOR. Thermal analysis studies were carried out to determine off gasing processes of ORNL AgIMOR. Independent of sample size, ~7wt% of total water is desorbed by 225°C. This includes both bulk surface and occluded water, and are monitored as H2O and OH. Of that total, ~5.5wt% is surface water which is removed by 125°C, and 1.5wt% is occluded (in zeolite pore) water. Less than ~1 wt% total water continues to desorb, but is completely removed by 500°C. Above 300°C, the detectable remaining desorbing species observed are iodine containing compounds, including I and I 2 .

09 BIOMASS FUELS↗
DOE OSTI2013

Development of low-pH cements for immobilisation of intermediate level radioactive waste: achievements and challenges

Although cementation is a widely recognized solidification/ stabilization process for immobilisation of Intermediate Level Radioactive Waste (ILRW), the low resistance to hyper-alkaline pore waters compromises the effectiveness of the process when Portland Cement (PC) is employed. Moreover the manufacture of PC is responsible for significant CO{sub 2} emissions. In this context, low pH cements are environmentally more suitable and have emerged as a potential alternative for obtaining secure waste forms. This paper summarises the achievements on development of low-pH cements and the challenges of using these new materials for the ILRW immobilisation. The performance of waste forms is also discussed in terms of radionuclides release. Reactive magnesium oxide and magnesium phosphate cements are emphasised as they feature important advantages such as consumption of available constituents for controlling acid-base reactions, reduced permeability and higher density. Additionally, in order to identify new opportunities for study, the long-term modelling approach is also briefly discussed. (authors)

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

Advanced Reactors Spent Fuel & Waste Science and Technology Program

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP ) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes three primary tasks the Spent Fuel & Waste Science and Technology (SFWST) Program is executing (or collaborating on) related to the back end of the nuclear fuel cycle (BENFC) for potential future advanced reactors: 1. Advanced Reactors Spent Fuel and Waste Streams: Strategies for the BENFC This set of activities define a high-level strategy for how to systematically approach, identify, and close research and development (R&D) gaps/potential issues associated with long-term management and final disposition of AR SNF and other possible AR waste streams. This task involves summarizing advanced reactor concepts, their likely SNF and other waste forms, and identifying previous experience with similar materials, for example from DOE-managed SNF, with closely related characteristics to the potential future AR SNF. Technical R&D gaps between the breadth of detailed understanding for safe storage, transportation and disposal of the existing light water reactor SNF fuel cycle (e.g., see NASEM, 2022) and potential future fuel cycles based on advanced reactors would then be identified. 2. Characterization and Packaging Options of Advanced Reactor SNF These activities evaluate characteristics and packaging options for advanced reactor spent fuel forms. The fuel forms are categorized into three types: (1) tri-structural isotropic (TRISO), (2) metallic, and (3) fuel salt. Emphasis is given to TRISO and metallic SNF and additional waste streams from such AR as driven by the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations1. Preliminary information for the spent-fuel salt discharged from molten-salt reactors (MSRs) is examined to provide a baseline for future efforts. All calculations and assumptions in this work are based on publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. 3. Back-End Management of Advanced Reactors (BEMAR) The DOE NE-8 organization has defined an Integrated Project Team to evaluate the Back End Management of Advanced Reactors (BEMAR) (includes DOE staff from a range of organizations (e.g., NE-81, NE-82, OCED) and national laboratory technical staff within the DOE NE-81 and NE-82 programs). This BEMAR group works directly with advance reactors developers to assess for the DOE the technical feasibility of storage, transportation, and disposal of AR SNF based on the characteristics provided by the developers to DOE (much of which is proprietary). The BEMAR is also tasked to develop rough-order-of-magnitude cost estimates to compare the waste management system for individual advanced reactors to existing light-water reactor management practices. To accomplish this, the BEMAR group is implementing a Systems Engineering approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Solid State Characterizations of Long-Term Leached Cast Stone Monoliths

This report describes the results from the solid phase characterization of six Cast Stone monoliths from the extended leach tests recently reported on (Serne et al. 2016),that were selected for characterization using multiple state-of-the-art approaches. The Cast Stone samples investigated were leached for > 590 d in the EPA Method 1315 test then archived for > 390 d in their final leachate. After reporting the long term leach behavior of the monoliths (containing radioactive 99 Tc and stable 127 I spikes and for original Westsik et al. 2013 fabricated monoliths, 238 U), it was suggested that physical changes to the waste forms and a depleting inventory of contaminants of potential concern may mean that effective diffusivity calculations past 63 d should not be used to accurately represent long-term waste form behavior. These novel investigations, in both length of leaching time and application of solid state techniques, provide an initial arsenal of techniques which can be utilized to perform such Cast Stone solid phase characterization work, which in turn can support upcoming performance assessment maintenance. The work was performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to characterize several properties of the long- term leached Cast Stone monolith samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

ETF Grout Expansion: Evaluation of MgO Additions and Struvite Precipitation Protocol for Potential Expansion

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 sulfate aqueous waste stream generated at the Hanford Effluent Treatment Facility (ETF). VSL did not observe expansion in samples prepared during the waste form development work. However, Savannah River National Laboratory (SRNL) personnel did encounter expansion, in some cases up to 20+ volume percent expansion in ETF grout samples while preforming work scope to evaluate compositional ranges for ETF brine and of the precipitation and solidification reagent. The 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. Additional evaluation was requested by WRPS SOW Requisitions #: 356730, Revision 1, February 1, 2022, to determine whether differences in the slags or differences 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 for grouting and curing 1L of Base Case ETF simulant. Materials were exchanged between VSL and SRNL and both laboratories performed the VSL struvite precipitation-solidification protocol and cured samples for at least 28 days. VSL and SRNL encountered expansion, at varying degrees, in all samples prepared under the material exchange and grouting protocol. Additional material exchange testing between VSL and SRNL and alternative materials evaluation to further investigate the causes of grout expansion and differentiate the effects of various dry materials, grouting procedures, and curing containers was requested by WRPS SOW Requisitions #: 356730, Revision 2, January 10, 2023. Pacific Northwest National Laboratory (PNNL) assembled and distributed materials to both laboratories for testing. Observations from this study indicated variations in MgO additions, water-to-dry mix (dm) ratios and curing containers did not correlate to ETF ammonia grout expansion. In comparison to previous work, SRNL observed the addition rate of NaOH to be a potential factor for the final pH in the struvite precipitation step. In the current testing NaOH was added in bulk (5-10 mL at one time) during the precipitation step. The resulting pH was 7.57-9.15 which is slightly higher than pH observed at a slower addition rate. Additional testing is needed to identify the presence and/or struvite stability in high-pH struvite precipitant ETF ammonia grout samples. Therefore, based on current data, it is unknown if NaOH addition rate is the fundamental cause of grout expansion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Effects of Iron Oxides on the Rheological Properties of Cementitious Slurry

Iron oxide has been considered a promising host for immobilizing and encapsulating radioactive 99Tc (t1/2=2.1x105 year), which significantly enhances the stability of 99Tc within a cementitious waste form. However, the flow behavior of cementitious slurry containing iron oxide has never been investigated to ensure its workability, which directly influences the preparation and performance of the cementitious waste form monolith. Variation in the rheological properties of the cementitious slurry were studied using rheometry and ultrasonic wave reflection to understand the effects of various iron oxides (magnetite, hematite, ferrihydrite, and goethite) during the cement setting and stiffening processes. The rheological behavior significantly varied with the addition of different chemical compounds of iron oxides. Complementary microscopic characteristics such as colloidal vibration currents, morphology, and particle size distributions further suggest that the most adverse alteration of cement setting and stiffening behavior caused by the presence of goethite may be attributed to its acicular shape.

cementitious slurry↗
DOE OSTI2013

Verification of the Accountability Method as a Means to Classify Radioactive Wastes Processed Using THOR Fluidized Bed Steam Reforming at the Studsvik Processing Facility in Erwin, Tennessee, USA - 13087

Studsviks' Processing Facility Erwin (SPFE) has been treating Low-Level Radioactive Waste using its patented THOR process for over 13 years. Studsvik has been mixing and processing wastes of the same waste classification but different chemical and isotopic characteristics for the full extent of this period as a general matter of operations. Studsvik utilizes the accountability method to track the movement of radionuclides from acceptance of waste, through processing, and finally in the classification of waste for disposal. Recently the NRC has proposed to revise the 1995 Branch Technical Position on Concentration Averaging and Encapsulation (1995 BTP on CA) with additional clarification (draft BTP on CA). The draft BTP on CA has paved the way for large scale blending of higher activity and lower activity waste to produce a single waste for the purpose of classification. With the onset of blending in the waste treatment industry, there is concern from the public and state regulators as to the robustness of the accountability method and the ability of processors to prevent the inclusion of hot spots in waste. To address these concerns and verify the accountability method as applied by the SPFE, as well as the SPFE's ability to control waste package classification, testing of actual waste packages was performed. Testing consisted of a comprehensive dose rate survey of a container of processed waste. Separately, the waste package was modeled chemically and radiologically. Comparing the observed and theoretical data demonstrated that actual dose rates were lower than, but consistent with, modeled dose rates. Moreover, the distribution of radioactivity confirms that the SPFE can produce a radiologically homogeneous waste form. The results of the study demonstrate: 1) the accountability method as applied by the SPFE is valid and produces expected results; 2) the SPFE can produce a radiologically homogeneous waste; and 3) the SPFE can effectively control the waste package classification. (authors)

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

Joint EM-NE-International Study of Glass Behavior over Geologic Time Scales

Vitrification has been chosen as the best demonstrated available technology for waste immobilization worldwide. To date, the contributions of physical and chemical processes controlling the long-term glass dissolution rate in geologic disposal remain uncertain; leading to a lack of international consensus on a glass corrosion rate law. Existing rate laws have overcome the uncertainty through conservatism, but a thorough mechanistic understanding of waste form durability in geologic environments would improve public and regulator confidence, as well as lead to cost savings if it is possible to take credit for the true durability of the waste form itself in system evaluations. To this end, six nations have joined together to formulate a joint plan for collaborative research into the mechanisms controlling the long-term corrosion of glass. This report highlights the technical program plan behind the US portion of this effort, with an emphasis on the current understanding (and limitations) of several mechanistic theories for glass corrosion. Some recent results are presented to provide an example of the ongoing research.

APT↗
DOE OSTI2023

Technology Development Roadmap for Volatile Radionuclide Capture and Immobilization

An "Off-Gas and Waste Forms Strategy Workshop" was held in Washington, D.C. on January 31, and February 1, 2023, to review the current state of the art (baseline) technologies for capturing and immobilizing these volatile radionuclides in waste forms (WFs). The discussions in that workshop were used to develop a roadmap for future research and development to mature technologies that are presently not ready for use in a commercial reprocessing facility and develop and demonstrate technologies that provide more safety, more simplicity, or lower costs compared to the current baseline.

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

Sulfur Polymer Stabilization/Solidification Treatability Study of Mercury Contaminated Soil from the Y-12 Site

As a result of past operations, the Department of Energy’s (DOE) Oak Ridge Y-12 National Security Complex (Y-12 Plant) has extensive mercury-contamination in building structures, soils, storm sewer sediments, and stream sediments, which are a source of pollution to the local ecosystem. Because of mercury’s toxicity and potential impacts on human health and the environment, DOE continues to investigate and implement projects to support the remediation of the Y-12 site. URS and #9122;CH2M Oak Ridge LLC (UCOR) under its prime contract with DOE has cleanup responsibilities on the DOE Oak Ridge Reservation and is investigating potential mercury-contaminated soil treatment technologies through an agreement with Babcock and Wilcox (B and W) Y-12, the Y-12 operating contractor to DOE. As part of its investigations, UCOR has subcontracted with Brookhaven National Laboratory (BNL) to conduct laboratory-scale studies evaluating the applicability of the Sulfur Polymer Stabilization/Solidification (SPSS) process using surrogate and actual mixed waste Y-12 soils containing mercury (Hg) at 135, 2,000, and 10,000 ppm. SPSS uses a thermoplastic sulfur binder to convert Hg to stable mercury sulfide (HgS) and solidifies the chemically stable product in a monolithic solid final waste form to reduce dispersion and permeability. Formulations containing 40 – 60 dry wt% Y-12 soil were fabricated and samples were prepared in triplicate for Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP) testing by an independent laboratory. Those containing 50 and 60 wt% soil easily met the study criteria for maximum allowable Hg concentrations (47 and 1 ppb, respectively compared with the TCLP limit of 200 ppb Hg). The lowest waste loading of 40 wt% yielded TCLP Hg concentrations slightly higher (240 ppb) than the allowable limit. Since the Y-12 soil tended to form clumps, the improved leaching at higher waste loadings was probably due to reduction in particle size from friction of the soil mixing, which creates more surface area for chemical conversion. This was corroborated by the fact that the same waste loading pre-treated by ball milling to reduce particle size prior to SPSS processing yielded TCLP concentrations almost 30 times lower, and at 8.5 ppb Hg was well below EPA limits. Pre-treatment by ball milling also allowed a reduction in the time required for stabilization, thus potentially reducing total process times by 30%.Additional performance testing was conducted including measurement of compressive strength to confirm mechanical integrity and immersion testing to determine the potential impacts of storage or disposal under saturated conditions. For both surrogate and actual Y-12 treated soils, waste form compressive strengths ranged between 2,300 and 6,500 psi, indicating very strong mechanical integrity (a minimum of greater than 40 times greater than the NRC guidance for low-level radioactive waste). In general, compressive strength increases with waste loading as the soil acts as an aggregate in the sulfur concrete waste forms. No statistically significant loss in strength was recorded for the 30 and 40 wt% surrogate waste samples and only a minor reduction in strength was measured for the 43 wt% waste forms. The 30 wt% Y-12 soil did not show a significant loss in strength but the 50 wt% samples were severely degraded in immersion due to swelling of the clay soil. The impact on Hg leaching, if any, was not determined.

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

Analysis of Hanford Cast Stone Supplemental LAW using Composition Adjusted SRS Tank 50 Salt Solution

Vitrification is the primary disposition path for Low Activity Waste (LAW) at the Department of Energy (DOE) Hanford Site. A cementitious waste form is one of the alternatives being considered for the supplemental immobilization of the LAW that will not be treated by the primary vitrification facility. Washington River Protection Solutions (WRPS) has been directed to generate and collect data on cementitious or pozzolanic waste forms such as Cast Stone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1991

Waste management technology development and demonstration programs at Brookhaven National Laboratory

Two thermoplastic processes for improved treatment of radioactive, hazardous, and mixed wastes were developed from bench scale through technology demonstration: polyethylene encapsulation and modified sulfur cement encapsulation. The steps required to bring technologies from the research and development stage through full scale implementation are described. Both systems result in durable waste forms that meet current Nuclear Regulatory Commission and Environmental Protection Agency regulatory criteria and provide significant improvements over conventional solidification systems such as hydraulic cement. For example, the polyethylene process can encapsulate up to 70 wt pct. nitrate salt, compared with a maximum of about 20 wt pct. for the best hydraulic cement formulation. Modified sulfur cement waste forms containing as much as 43 wt pct. incinerator fly ash were formulated, whereas the maximum quantity of this waste in hydraulic cement is 16 wt pct.

Kalb, Paul D.↗
DOE OSTI2013

Hanford's Simulated Low Activity Waste Cast Stone Processing

Cast Stone is undergoing evaluation as the supplemental treatment technology for Hanford’s (Washington) high activity waste (HAW) and low activity waste (LAW). This report will only cover the LAW Cast Stone. The programs used for this simulated Cast Stone were gradient density change, compressive strength, and salt waste form phase identification. Gradient density changes show a favorable outcome by showing uniformity even though it was hypothesized differently. Compressive strength exceeded the minimum strength required by Hanford and greater compressive strength increase seen between the uses of different salt solution The salt waste form phase is still an ongoing process as this time and could not be concluded.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Recent Advances in Corrosion Science Applicable To Disposal of High-Level Nuclear Waste

High level radioactive waste is accumulating at temporary storage locations around the world and will eventually be placed in deep geological repositories. Many different types of geological formation are under consideration, but all will eventually allow water to contact the metallic waste containers considering that the period of performance required to allow sufficient decay of dangerous radionuclides is on the order of 105 – 106 years. Corrosion of the containers and waste forms in the aqueous repository environment is therefore a concern. This review describes the recent advances of the field of materials corrosion based on the work of the Center for Performance and Design of Nuclear Waste Forms and Containers (WastePD) to address the issues associated with the long-term performance assessment and the design of materials with improved performance, where performance is defined as resistance to aqueous corrosion. Glass, crystalline ceramics, and metals are discussed separately, and then the near field interactions of these different materials classes are addressed. Recommendations for future directions are provided.

Frankel, Gerald↗
DOE OSTI2012

The Aqueous Corrosion Behavior of Technetium – Iron Alloy Materials

If reprocessed, one by-product of spent nuclear fuel is technetium. Metal waste forms are currently being considered for the long-term storage of technetium. Technetium may be alloyed with stainless steel in order to enhance processing. As an initial step in understanding the corrosion behavior of these waste forms, the corrosion of binary technetium alloys has been studied. Technetium was alloyed with iron across a broad range of compositions. It was found that both chemical and microstructural effects play a role in corrosion behavior with corrosion minimized at intermediate technetium concentrations. Preferential dissolution of iron was found to occur at the extremes of the concentration ranges.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Assessment of Methods to Consolidate Iodine-Loaded Silver-Functionalized Silica Aerogel

The U.S. Department of Energy is currently investigating alternative sorbents for the removal and immobilization of radioiodine from the gas streams in a nuclear fuel reprocessing plant. One of these new sorbents, Ag 0 -functionalized silica aerogels, shows great promise as a potential replacement for Ag-bearing mordenites because of its high selectivity and sorption capacity for iodine. Moreover, a feasible consolidation of iodine-loaded Ag0-functionalized silica aerogels to a durable SiO 2 -based waste form makes this aerogel an attractive choice for sequestering radioiodine. This report provides a preliminary assessment of the methods that can be used to consolidate iodine-loaded Ag 0 -functionalized silica aerogels into a final waste form. In particular, it focuses on experimental investigation of densification of as prepared Ag 0 - functionalized silica aerogels powders, with or without organic moiety and with or without sintering additive (colloidal silica), with three commercially available techniques: 1) hot uniaxial pressing (HUP), 2) hot isostatic pressing (HIP), and 3) spark plasma sintering (SPS). The densified products were evaluated with helium gas pycnometer for apparent density, with the Archimedes method for apparent density and open porosity, and with high-resolution scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) for the extent of densification and distribution of individual elements. The preliminary investigation of HUP, HIP, and SPS showed that these sintering methods can effectively consolidate powders of Ag 0 -functionalized silica aerogel into products of near-theoretical density. Also, removal of organic moiety and adding 5.6 mass% of colloidal silica to Ag0-functionalized silica aerogel powders before processing provided denser products. Furthermore, the ram travel data for SPS indicated that rapid consolidation of powders can be performed at temperatures below 950°C.

Ag-Functionalized Silica Aerogel↗