Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Waste Forms”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Radiation stability test on multiphase glass ceramic and crystalline ceramic waste forms

A radiation stability study was performed on glass ceramic and crystalline ceramic waste forms. These materials are candidate host materials for immobilizing alkali/alkaline earth (Cs/Sr-CS) + lanthanide (LN) + transition metal (TM) fission product waste streams from nuclear fuel reprocessing. In this study, glass ceramics were fabricated using a borosilicate glass as a matrix in which to incorporate CS/LN/TM combined waste streams. The major phases in these multiphase materials are powellite, oxyaptite, pollucite, celsian, and durable residual glass phases. Al2O3 and TiO2 were combined with these waste components to produce multiphase crystalline ceramics containing hollandite-type phases, perovskites, pyrochlores and other minor metal titanate phases. For the radiation stability test, selected glass ceramic and crystalline ceramic samples were exposed to different irradiation environments including low fluxes of high-energy (~1–5 MeV) protons and alpha particles generated by an ion accelerator, high fluxes of low-energy (hundreds of keV) krypton particles generated by an ion implanter, and in-situ electron irradiations in a transmission electron microscope. These irradiation experiments were performed to simulate self-radiation effects in a waste form. Ion irradiation-induced microstructural modifications were examined using X-ray diffraction and transmission electron microscopy. Our preliminary results reveal different radiation tolerance in different crystalline phases under various radiation damage environments. However, their stability may be rate dependent which may limit the waste loading that can be achieved.

Crystalline ceramic↗
DOE OSTI2014

Ceramic Waste Form Data Package

The purpose of this data package is to provide information about simulated crystalline waste forms that can be used to select an appropriate composition for a Cold Crucible Induction Melter (CCIM) proof of principle demonstration. Melt processing, viscosity, electrical conductivity, and thermal analysis information was collected to assess the ability of two potential candidate ceramic compositions to be processed in the Idaho National Laboratory (INL) CCIM and to guide processing parameters for the CCIM operation. Given uncertainties in the CCIM capabilities to reach certain temperatures throughout the system, one waste form designated 'Fe-MP' was designed towards enabling processing and another, designated 'CAF-5%TM-MP' was designed towards optimized microstructure. Melt processing studies confirmed both compositions could be poured from a crucible at 1600°C although the CAF-5%TM-MP composition froze before pouring was complete due to rapid crystallization (upon cooling). X-ray diffraction measurements confirmed the crystalline nature and phase assemblages of the compositions. The kinetics of melting and crystallization appeared to vary significantly between the compositions. Impedance spectroscopy results indicated the electrical conductivity is acceptable with respect to processing in the CCIM. The success of processing either ceramic composition will depend on the thermal profiles throughout the CCIM. In particular, the working temperature of the pour spout relative to the bulk melter which can approach 1700°C. The Fe-MP composition is recommended to demonstrate proof of principle for crystalline simulated waste forms considering the current configuration of INL's CCIM. If proposed modifications to the CCIM can maintain a nominal temperature of 1600°C throughout the melter, drain, and pour spout, then the CAF-5%TM-MP composition should be considered for a proof of principle demonstration.

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

Development and Durability Testing of a Low‐Temperature Sintering Bi–Si–Zn Oxide Glass Composite Material (GCM) 129 I Waste Form

The capture and safe storage of radiological iodine ( 129 I) from nuclear fuel reprocessing is of concern due to its long half‐life and potential mobility in the environment. The development of durable waste forms in which to store captured iodine requires materials that are both compatible with the iodine capture phases and durable to repository environments. To that end, Sandia is developing a low‐temperature sintering Bi–Si–Zn oxide glass composite material (GCM) waste form and herein presents results of durability testing. Furthermore, durability studies were extended to both the occluded iodine capture material Ag‐Zeolite (Mordenite, MOR) as well as the GCM, synthesized with compositional variations including: amount of Ag flake added, AgI‐MOR particle size in the GCM, and mass loading of I within the AgI‐MOR. Product consistency test (PCT‐B), chemical durability (MCC‐1) tests, and single‐pass flow‐through (SPFT) tests, were performed on both the individual components of the GCM and the completed GCMs. Durability tests indicate low GCM dissolution rates (<10 −3 g/m 2 ·d) across wide variable ranges including: pH, AgI‐MOR loading, I loading, and AgI‐MOR particle size. Results indicate that the Bi–Si–Zn oxide glass matrix sharply limits the release of iodine from the otherwise relatively fast degrading AgI‐MOR getter material. Furthermore, the formation of an amorphous AgI phase during sintering of the GCM results in the limitation of iodine release during waste form degradation. Durability of GCM and release rates approximate those of established nuclear waste glasses, or analogues such as basaltic glass. This suggests that the Bi–Si–Zn GCM is a viable candidate as a repository iodine waste form.

Mowry, Curtis D.↗
DOE OSTI2021

The role of environmental conditions on the carbonation of an alkali-activated cementitious waste form

Cast Stone is a cementitious waste form being considered for the solidification of low activity waste at the DOE Hanford Site. Under near-surface disposal conditions in an arid environment, Cast Stone is subject to drying and carbonation which may impact retention of waste constituents. Here this study investigates the effects of environmental CO 2 concentration and relative humidity (RH) on the carbonation of Cast Stone. The rate of carbonation front ingress and the extent of carbonation reaction were characterized for samples aged up to 48 weeks at three RH levels and two CO 2 concentrations. While the 68% RH environment allowed the greatest reaction extent, the 15% RH environment yielded the deepest carbonation front ingress. At 68% RH, there was a linear relationship between the ingress rate and CO 2 concentration. Carbonation reactions increased the drying rate of Cast Stone. Redistribution of sodium toward the wetting front was observed under drying and carbonation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Glass-Bonded Monazite Waste Forms for Lanthanide and Actinide Immobilization: From Theoretical Design to Scale-Up Production and Characterization

The development of nuclear waste forms for both existing and future nuclear wastes is critical to ensuring global environmental safety. This study focuses on waste management from molten salt reactors, where fuel exists in a salt form and could be processed in real time for the removal of neutron poisons such as xenon isotopes (e.g., 135 Xe) and rare earth elements (REEs, e.g., 149 Sm). To ensure safe, stable, and long-term disposal in geological repositories, REEs must be incorporated into a durable waste form. Iron-phosphate glasses are a promising candidate due to their low melting points, high chemical durability, and their ability to incorporate high concentrations of REEs. In this study, we successfully prepared iron-phosphate glass waste forms with high Nd loadings (up to 37 mass %) in batch sizes ranging from small (23 g) to large (1600 g). The resulting materials contained up to 75 mass % NdPO 4 , contributing to their mechanical resilience and exceptional chemical durability. These findings highlight the potential of iron-phosphate glasses as high-efficiency, chemically durable waste forms and demonstrate the successful transition from theoretical design to scaled-up production.

amorphous materials↗
DOE OSTI2021

Evaluation of Degradation Mechanisms for Solid Secondary Waste Grout Waste Forms

The overall objective of this work is to provide defensibility for the long-term performance of grouted Hanford SSW streams when disposed in a near surface disposal facility, IDF, on the Hanford Site. Providing defensibly for the long-term performance of the grouted waste forms is consistent with the research and development activities identified in the Performance Assessment Maintenance Plan, (Westcott et al. 2019), that are necessary to address the assumptions made in the PA. Specifically, this work addresses two areas identified for further research and development activities in the plan: (1) “Evaluate ongoing research on transport characteristics of cementitious materials using accelerated tests to approximate the effects of aging/alteration/weathering”; and (2) “Evaluate ongoing research on microbial effects on transport processes in cementitious materials.” The assembled subject matter expert team evaluated a list of degradation mechanisms and supporting processes and provided rankings of areas where further research and development (R&D) are needed. From this assessment, high priority R&D areas include: (1) the effects of carbonation, Ca leaching, and SSW dimensional change in grout waste forms; (2) updated model representations of grouted waste forms; and (3) scaled testing demonstrations. Moderate priority items including a paper study on possible microbial influence, reoxidation rates, radionuclide/contaminant dissolution from SSW in the grout and freeze thaw behavior. Other processes evaluated were either deemed unlikely to occur, to have little impact on the SSW waste forms, or to occur at time frames beyond those considered (>10,000 years). The assessments and proposed R&D approaches are expected to support an update to the WRPS SSW roadmap.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Reassessment of Dose Contributions from the Paducah Cask Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater and inadvertent human intruder screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in the addition of U-235 being brought back into the list of radionuclides requiring inventory limits within the Paducah Cask special waste form (SWF). In addition, while reviewing the earlier vadose zone transport analysis efforts, an inconsistency was observed in the initial U-235P inventory that was supplied to the Consolidated Waste Tracking System (CWTS) back in 2009.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Secondary Waste Form Down-Selection Data Package—DuraLith

This data package developed for the DuraLith wasteform includes information available in the open literature and from data obtained from testing currently underway. DuraLith is an alkali-activated geopolymer waste form developed by the Vitreous State Laboratory at The Catholic University of America (VSL-CUA) for encapsulating liquid radioactive waste. A DuraLith waste form developed for treating Hanford secondary waste liquids is prepared by alkali-activation of a mixture of ground blast furnace slag and metakaolinite with sand used as a filler material. Based on optimization tests, solid waste loading of {approx}7.5% and {approx}14.7 % has been achieved using the Hanford secondary waste S1 and S4 simulants, respectively. The Na loading in both cases is equivalent to {approx}6 M. Some of the critical parameters for the DuraLith process include, hydrogen generation and heat evolution during activator solution preparation using the waste simulant, heat evolution during and after mixing the activator solution with the dry ingredients, and a working window of {approx}20 minutes to complete the pouring of the DuraLith mixture into molds. Results of the most recent testing indicated that the working window can be extended to {approx}30 minutes if 75 wt% of the binder components, namely, blast furnace slag and metakaolin are replaced by Class F fly ash. A preliminary DuraLith process flow sheet developed by VSL-CUA for processing Hanford secondary waste indicated that 10 to 22 waste monoliths (each 48 ft3 in volume) can be produced per day. There are no current pilot-scale or full-scale DuraLith plants under construction or in operation; therefore, the cost of DuraLith production is unknown. The results of the non-regulatory leach tests, EPA Draft 1313 and 1316, Waste Simulant S1-optimized DuraLith specimens indicated that the concentrations of RCRA metals (Ag, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the EPA draft 1315 leach test showed that LI values for COCs, namely 99Tc and I, ranged from 8.2 to 11.4 and 4.3 to 7.5, respectively. These values indicate that 99Tc meets the WAC LI requirement of 9.0 whereas, the LI values for I does not meet the WAC requirement of 11.0. Results of Toxicity Characteristic Leaching Procedure (TCLP)(EPA Method 1311) conducted on Waste Simulant S1-optimized DuraLith specimens, indicated that the concentrations of RCRA metals (Ag, As, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268.48. The data from the ANSI/ANS 16.1 leach test showed that LI values for COC, namely Re (as a Tc surrogate), ranged from 8.06 to 10.81. The LI value for another COC, namely I, was not measured in this test. The results of the compressive strength testing of Waste Simulant S1-optimized DuraLith specimens indicated that the monoliths were physically robust with compressive strengths ranging from 115.5 MPa (16757 psi) to 156.2 MPA (22667 psi).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Investigation of processing advanced waste forms in Cold Crucible Induction Melter systems

The Cold Crucible Induction Melter (CCIM) technology has been identified in several independent reviews and studies as a potential alternative or supplemental vitrification technology for specific U.S. Department of Energy (DOE) high level radioactive waste (HLW) streams. These inventories have unique chemistries, such as high alumina, high iron, high chromium, etc., that are not amenable to efficient immobilization in conventional borosilicate glass (BSG) matrices. Advanced waste forms, such as iron phosphate glass (FePG), high crystalline BSG (HC-BSG), and alumino-silicate glass-ceramic (GC) have been shown to provide significantly improved immobilization matrices for these challenging waste streams, including dramatically improved waste loadings; however, they are difficult to efficiently process in the baseline joule-heated ceramic melters (JHCMs). The CCIM technology can effectively process these advanced waste forms at the desired high waste loadings. This paper describes the most recent collaborative activities between the DOE Office of Environmental Management (DOE EM) and various Russian institutions in modeling, development and testing of advanced waste forms in CCIMs, including FePG and HC-BSG formulations. Additionally, past interactions with other international partners, including France and Korea, are described. Finally, recent enhancements to CCIM processing related to modeling, draining techniques and automated control, which were collaboratively developed, will be discussed. (authors)

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

Optimization studies on GCM for iodine waste forms

We are purusing an understand of the durability and materials processability of the low temperature sintering Bi-Si oxide Glass Composite Material (GCM)1 Waste Form for iodine capture materials. The chemical and physical controls over iodine release from candidate 129 I waste forms must be quantified to predict long-term waste form effectiveness.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Cs 3 Bi 2 I 9 -hydroxyapatite composite waste forms for cesium and iodine immobilization

Perovskite-based ceramic composites were developed as potential waste form materials for immobilizing cesium (Cs) and iodine (I) with high waste loadings and chemical durability. The perovskite Cs 3 Bi 2 I 9 has high Cs (22 wt%) and I (58 wt%) content, and thus can be used as a potential host phase to immobilize these critical radionuclides. In this work, the perovskite Cs 3 Bi 2 I 9 phase was synthesized by a cost effective solution-based approach, and was embedded into a highly durable hydroxyapatite matrix by spark plasma sintering to form dense ceramic composite waste forms. The chemical durabilities of the monolithic Cs 3 Bi 2 I 9 and Cs 3 Bi 2 I 9 -hydroxyapatite composite pellets were investigated by static and semi-dynamic leaching tests, respectively. Cs and I are incongruently released from the matrix for both pure Cs 3 Bi 2 I 9 and composite structures. The normalized Cs release rate is faster than that of I, which can be explained by the difference in the strengths between Cs-I and Bi-I bonds as well as the formation of insoluble micrometer-sized BiOI precipitates. The activation energies of elemental releases based on dissolution and diffusion-controlled mechanisms are determined with significantly higher energy barriers for dissolution from the composite versus that of the monolithic Cs 3 Bi 2 I 9 . The ceramic-based composite waste forms exhibit excellent chemical durabilities and waste loadings, commensurate with the state-of-the-art glass-bonded perovskite composites for I and Cs immobilization.

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

Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

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

Radiation damage of hollandite in multiphase ceramic waste forms

Radiation damage was simulated in multiphase titanate-based ceramic waste forms using an ion accelerator to generate high energy alpha particles (He + ) and an ion implanter to generate 7 MeV gold (Au 3+ ) particles. X-ray diffraction and transmission electron microscopy were used to characterize the damaged surfaces and nearby regions. Simulated multiphase ceramic waste forms were prepared using two processing methods: spark plasma sintering and melt-processing. Both processing methods produced ceramics with similar phase assemblages consisting of hollandite-, zirconolite/pyrochlore-, and perovskite- type phases. The measured heavy ion (Au 3+ ) penetration depth was less in spark plasma sintered samples than in melt-processed samples. Furthermore, structural breakdown of the hollandite phase occurred under He + irradiation indicated by the presence of x-ray diffraction peaks belonging to TiO 2 , BaTiO 5 , and other hollandite related phases (Ba 2 Ti 9 O 20 ). The composition of the constituent hollandite phase affected the extent of damage induced by Au 3+ ions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A low-temperature, one-step synthesis for monazite can transform monazite into a readily usable advanced nuclear waste form

It has been demonstrated that monazite-type materials are excellent candidates for nuclear waste forms, and hence, their facile synthesis is of great importance for the needed sequestration of existing nuclear waste. The synthesis of monazite, LaPO 4 , requires inconveniently high temperatures near 1000°C and generally involves the conversion of the presynthesized rhabdophane, LaPO 4 •nH 2 O, to the LaPO 4 monazite phase. During this structure transformation, the rhabdophane converts irreversibly to the thermodynamically stable monoclinic monazite structure. A low-temperature (185° to 260°C) mild hydrothermal acid-promoted synthesis of monazite is described that can both transform presynthesized rhabdophane or assemble reagents to the monoclinic monazite structure. The pH dependence of this reaction is detailed, and its applicability to the LnPO 4 (Ln = La, Ce, Pr, Nd, Sm-Gd), Ca 0.5 Th 0.5 PO 4 , and Sr 0.5 Th 0.5 PO 4 systems is discussed. The crystal growth of Ca 0.5 Th 0.5 PO 4 and Sr 0.5 Th 0.5 PO 4 is described, and their crystal structures were reported. In situ x-ray diffraction studies, performed as a function of temperature, provide insight into the structure transformation process.

Biogeochemistry↗
DOE OSTI2020

Novel complex ceramic oxides, Ln 2 TiO 5 (Ln = La, Sm, Gd, Tb, Dy, Ho, Er, and Yb), for polyphase nuclear waste‐forms

Abstract As part of a broader study of ceramic nuclear waste‐forms, four different lanthanide titanates were fabricated; La 0.1 Sm 0.1 Gd 0.1 Tb 0.1 Dy 0.3 Ho 0.1 Er 0.2 YbTiO 5 , Sm 0.3 Gd 0.3 Dy 0.3 Yb 1.1 TiO 5 , Sm 0.1 Gd 0.4 Dy 0.4 Yb 1.1 TiO 5 , and Sm 0.2 Gd 0.2 Dy 0.2 Yb 1.4 TiO 5 . The aim was to produce single‐phase novel materials with cubic symmetry, capable of incorporating a wide variety of cations and with acceptable radiation tolerance. The chemistry flexibility and radiation tolerance are some of the major desirable properties for nuclear waste‐form materials. By using multiple lanthanides the average lanthanide radius can be controlled and consequently the structure, along with properties such as radiation tolerance. The radiation tolerance was assessed using in situ 1 MeV krypton irradiation and transmission electron microscopy characterization. Those materials for which cubic symmetry was achieved displayed better radiation tolerance; a greater critical fluence of ions ( F c ) was required for the crystalline to amorphous transition, and a lower temperature was required to maintain crystallinity ( T c ) during irradiation.

Aughterson, Robert D.↗
DOE OSTI2020

Glass waste form options for rare-earth fission products from electrochemical reprocessing

Herein, we describe a first-of-a-kind study where lead tellurite (Pb-Te-O), iron phosphate (Fe-P-O), and lanthanide aluminoborosilicate (LABS) waste forms are evaluated for treating and/or immobilizing waste streams containing high rare-earth (RE) content, e.g., RECl 3 , REOCl, and/or REO x . Results from this study show that NdOCl can be immobilized in 78TeO 2 -22PbO glass at up to 10 mass%. Higher loadings of up to 15 mass% were demonstrated, but with a small fraction of crystallization. Two Fe-P-O glasses produced from either RECl 3 (RE = Nd,Ce) or NdOCl revealed products with very similar properties, demonstrating that the phosphate process can be used to chlorinate a variety of RE-containing waste streams. Two LABS glasses produced from Nd 2 O 3 or NdOCl showed similar properties, but the LABS-NdOCl glass did have some residual Cl present. Characterization is provided for all waste forms including chemical analysis, X-ray diffraction, and bulk density.

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

X-ray diffraction of slag-based sodium salt waste forms

The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.

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

Corrosion interactions between stainless steel and lead vanado-iodoapatite nuclear waste form part I

This study explores the corrosion interactions between a metallic canister material, stainless steel (SS) 316, and an I-bearing ceramic waste form, lead vanado-iodoapatite (I-APT, Pb 9.85 (VO 4 ) 6 I 1.7 ), in a chloride solution. Crevice corrosion of the SS in close proximity to the I-APT resulted in the development of an aggressive environment at the interface of the two materials, which was acidic and enriched in Cl – anions. I-APT also corroded in the crevice region, primarily through ion-exchange between the I – ions from the I-APT matrix and anions from the environment. The enrichment of Cl – anions within the occluded crevice space as the result of SS crevice corrosion enhanced the corrosion of I-APT. The release of iodine from this apatite waste form could be accelerated owing to this mechanism. This is evidenced by a depletion of iodine from the I-APT matrix and a large amount of Cl-bearing precipitates on the surfaces of both SS and I-APT. On the other hand, the corrosion of I-APT leads to the precipitation of a V- and Pb-rich layer, which inhibits the localized corrosion of SS to an extent. This study advances the understanding of the near-field corrosion interactions between metallic canisters and ceramic waste forms.

36 MATERIALS SCIENCE↗