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.

SourceResearch connectionFrom yearThrough year

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

At least 217 records · Page 12

All sources NASA NTRS DOE OSTI

DOE OSTI2026

Structural Origins of High MoO 3 Solubility in Peraluminous Borosilicate Glasses

Molybdenum (Mo) imposes strict loading limits in conventional borosilicate nuclear waste glasses due to the tendency of tetrahedral molybdate [MoO 4 ] 2− species to phase-separate and crystallize as alkali molybdates. Here, we demonstrate an unprecedented 13.96 wt % (7.51 mol %) MoO 3 solubility in peraluminous sodium aluminoborosilicate glasses a ∼15× increase over their peralkaline counterparts. Using Raman spectroscopy, multinuclear and dipolarcorrelation magic angle spinning nuclear magnetic resonance (MAS NMR), electron paramagnetic resonance (EPR), and scanning transmission electron microscopy (STEM)-energy dispersive spectroscopy (EDS), we reveal that Nadeficient, low optical basicity conditions stabilize octahedral MoO 6 units, which polymerize into molybdite-like Mo−O clusters dispersed within the glass matrix. These Mo-rich clusters suppress the formation of depolymerized [MoO 4 ] 2− environments typically responsible for Na 2 MoO 4 precipitation and instead promote the formation of Na 2 Mo 2 O 7 as the saturation phase. Concurrently, Mo solubility drives the conversion of AlO 4 − to higher-coordination AlO 5 species, liberating Na + that is subsequently sequestered in molybdate-rich domains. The combined evolution of Mo coordination, modifier redistribution, and network depolymerization provides a mechanistic basis for the markedly enhanced Mo solubility in peraluminous compositions. These findings establish new structural guidelines for designing aluminoborosilicate waste forms with substantially greater capacity to incorporate Mo-rich nuclear waste streams.

Amorphous materials↗

DOE OSTI2020

Reactions during conversion of simplified low-activity waste glass feeds

The mechanisms that affect the incorporation of 99Tc, a volatile radioactive component of concern, into glass melt during vitrification of low-activity waste (LAW) are being investigated to develop the method to increase the retention of 99Tc in glass waste form. Previous studies with simulated LAW glass feeds (slurry mixture of liquid waste and chemical/mineral additives) demonstrated that the early stage feed-to-glass conversion reactions below 800°C are critical for the Re (used as a nonradioactive surrogate of 99Tc) retention in glass. To examine the effect of feed composition on the feed-to-glass conversion reactions, simplified systems containing major LAW components (NaNO3 and NaOH) and representative additive components (SiO2 and H3BO3) were designed and tested. The ratio of H3BO3 to NaNO3 was varied in three-component system without NaOH and that of NaOH to NaNO3 was varied in the four-component system at a fixed H3BO3 to NaNO3 ratio. As a first step of testing with simplified feeds, this study applied thermal analyses and phase characterization of the reacting feeds, which were performed without the addition of Re, to investigate the evolution of salt phases during slurry drying process and upon heating of dried feeds.

Jin, Tongan↗

DOE OSTI2022

Follow-on Report of Analysis of Approaches to Supplemental Treatment of Low-Activity Waste at the Hanford Nuclear Reservation (Vol. I)

The Hanford Site, in southeast Washington State, is preparing to disposition approximately 56,000,000 gallons (56 Mgal) of radioactive and chemically hazardous wastes currently stored in underground tanks at the site. Tank wastes will be divided into a high-activity fraction and a low-activity fraction for subsequent treatment and disposition. A waste processing and treatment facility, the Waste Treatment and Immobilization Plant (WTP), will include the high-level waste (HLW) vitrification facility (WTP HLW Vitrification Facility) for immobilizing the high-activity fraction and a low-activity waste (LAW) vitrification facility (WTP LAW Vitrification Facility) for immobilizing the low-activity fraction. Both facilities will use vitrification technology to immobilize the Hanford tank wastes in a glass waste form. The volume of LAW to be treated and disposed of following waste retrieval and WTP operations will exceed the planned processing capacity of the WTP LAW Vitrification Facility. ORP-11242,-River Protection Project System Plan, estimates a shortfall in LAW treatment capacity of approximately 56 Mgal, approximately 50% of the projected LAW volume. To maintain the planned tank waste processing mission schedule, the U.S. Department of Energy (DOE) will require additional LAW treatment capacity (termed “supplemental LAW”) external to the WTP process. LAW must be solidified by a treatment technology before the waste can be permanently disposed of in an approved DOE on-site disposal facility or a commercial (state or U.S. Nuclear Regulatory Commission [NRC-licensed]) off-site mixed low-level waste disposal facility. A decision on the approach to supplemental LAW treatment, processing, and disposal has not yet been made

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Solidification of REDC Organics For Disposal As Solid Waste

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the REDC organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid waste. Solidifying the organic liquids using PM-199 Organoclay ® would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours to produce a solid wasteform.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Structures of fluoride containing aluminosilicate low activity nuclear waste glasses: A molecular dynamics simulations study

Fluorine and other halides commonly exist in nuclear waste forms, and due to their volatile nature, halide retention poses an issue affecting waste loading during vitrification. The compositional effect on fluorine incorporation in aluminosilicate glasses is investigated through molecular dynamics simulations. Oxygen and fluorine coordination numbers around glass former and modifier cations, bond angle distributions, and medium range structure features such as Q n distributions, ring size distributions and neutron diffraction structure factors were calculated. It was found that fluorine has higher preference to bond to Ca 2+ than to Na + , both in the melt and the glass, and there is no Si-F bond formation in the glass but they do exist in the melt. Consequently, CaO for Na 2 O substitution can be an effective way to help fluorine retention without significantly changing the glass chemistry. Furthermore, these results thus provide insights on fluorine incorporation in the aluminosilicate nuclear waste glasses and the strategy on how to improve fluorine retention both in the glass and the melt.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Vibrational spectroscopy of uranium tetrafluoride hydrates

Uranium tetrafluoride (UF 4 ) is an important intermediate in the production of UF6 and nuclear fuel. Historical characterization of UF 4 with Raman spectroscopy was plagued with ambiguity until the first accurate Raman spectrum was published by our group in 2016. Although generally considered to be relatively stable, UF 4 can hydrolyze to form numerous UF 4 hydrates that may play a role in future uranium waste forms. In contrast to anhydrous UF 4 , the hydrates, with their OH stretch and HOH bending modes, can be spectroscopically characterized by the type and degree of water bonding in the crystal lattice, which can yield additional information about their crystal structure. Herein, vibrational spectroscopy (Raman and infrared) was used to characterize three different UF 4 hydrates: UF 4 (H 2 O) 0.33 , U 3 F 12 (H 2 O), and UF 4 (H 2 O) 2.5 . Spectra show the different hydrates vary in the number of observed bands, full-width half-maximum of the bands, and band intensity. These differences are due to varying interactions between the OH stretch and HOH bending modes with UF 4 and the polymeric UF 4 structure in the crystal lattice. These vibrational data, in combination with spectral fitting and crystallographic structures measured with powder X-ray diffraction and single crystal X-ray diffraction, provide unique details on the location of water molecules in the crystal lattice of hydrated UF 4 , and provide an interesting contrast to the vibrational spectra of anhydrous UF 4 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Follow-On Report of Analysis of Approaches to Supplemental Treatment of Low-Activity Waste at the Hanford Nuclear Reservation (Volumes I & II)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Demonstration Test Report: Vacuum Thermal Desorption (VTD) Testing to Support West Area Waste Feed Test 2

Approximately 54 million gallons of radioactive and hazardous waste are stored in underground storage tanks at the U.S. Department of Energy’s (DOE’s) Hanford Site in southeastern Washington State. The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to immobilize the waste in glass waste forms. The baseline configuration of the WTP includes a Pretreatment Facility to separate the wastes into a small volume of high-level waste (HLW) containing most of the radioactivity and a larger volume of low-activity waste (LAW) containing most of the non-radioactive chemicals. The HLW will be converted to glass in the HLW vitrification facility for ultimate disposal at a federal repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Summary Report For The Analysis Of The Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample For Canister S04023 (Rev. 1)

In order to comply with the Defense Waste Processing Facility (DWPF) Waste Form Compliance Plan for Sluldge Batch 7b, Savannah River National Laboratory (SRNL) personnel characterized the Defense Waste Processing Facility (DWPF) pour stream (PS) glass sample collected while filling canister S04023. This report summarizes the results of the compositional analysis for reportable oxides and radionuclides and the normalized Product Consistency Test (PCT) results. The PCT responses indicate that the DWPF produced glass that is significantly more durable than the Environmental Assessment glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Summary Report for the Analysis of the Sludge Batch 7b (Macrobatch 9) DWPF Pour Stream Glass Sample for Canister S04023

In order to comply with the Defense Waste Processing Facility (DWPF) Waste Form Compliance Plan for Sludge Batch 7b, Savannah River National Laboratory (SRNL) personnel characterized the Defense Waste Processing Facility (DWPF) pour stream (PS) glass sample collected while filling canister S04023. This report summarizes the results of the compositional analysis for reportable oxides and radionuclides and the normalized Product Consistency Test (PCT) results. The PCT responses indicate that the DWPF produced glass that is significantly more durable than the Environmental Assessment glass. Results and further details are documented in SRNL-STI-2013-00462, Revision 1 “Analysis of the DWPF Sludge Batch 7b (Macrobatch 9) Pour Stream Glass Sample.”

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1977

An analysis of the technical status of high level radioactive waste and spent fuel management systems

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

English, T.↗

DOE OSTI2022

OPP-B332-032 Rev AD

The purpose of this procedure is to provide the approved method for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

OPP-B332-035 Rev AD

This procedure provides instructions for converting transuranic (TRU) contaminated liquid waste into acceptable TRU waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Insights on the structure and properties of sodium iron phosphate glasses from molecular dynamics simulations

Iron phosphate glasses are promising nuclear waste forms while more detailed understanding of their structures and structure-property relations are still needed to better design waste glass compositions. In this work we report studies of three series of sodium iron phosphate (NFP) glasses: 60P 2 O 5 -(40-x)Fe 2 O 3 -xNa 2 O (x = 0→40), (100–2x)P 2 O 5 -xFe 2 O 3 -xNa 2 O (x = 5→17.5) and one with different iron redox ratio, to understand the composition as well as the iron redox effects on the structure and properties of these glasses using molecular dynamics simulations with effective two-body and three-body potentials. Structural analyses, including pair distribution function, bond angle distribution, Q n distribution, and polyhedral connectivity, were performed to obtain in-depth information on short-range and medium-range structural features. The P-O pair distributions showed a first peak splitting with phosphorus-bridging and non-bridging oxygen contributions. This and the average P-O and other cation-oxygen bond distances are in excellent agreement with experiments. The coordination number of P 5+ remained four while that of Fe 3+ increased from 4.30 to 4.72 with decreasing Fe/Na ratio. Polyhedral linkage analysis showed [PO 4 ] units linked with [PO 4 ] and [FeO x ] through corner-sharing while the [PO 4 ]-[FeO x ] linkages become dominant for compositions with Fe 2 O 3 larger than 15 mol%. The effect of iron redox ratio on the structure of NFP glasses was also studied and it was found that bond lengths and coordination numbers were not strongly affected, while the reduction of iron introduced higher network distortions, as evident by O-P-O bond angle and Q n distribution. The glass transition temperature (T g ) showed a monotonic increase with Fe 2 O 3 in the first series, in good agreement with experiments, while those of the second series showed a maximum at P 2 O 5 = 82 mol%. Here, calculated elastic moduli were found to increase with Fe 2 O 3 in the first glass series, which was be explained by the increase of network connectivity, while those of the second series decrease with Fe 2 O 3 due to decrease of P 2 O 5 .

36 MATERIALS SCIENCE↗

DOE OSTI2021

Glass Crystallization Constraints for WTP LAW Operations: Assessment of Isothermal Treatments on Crystal Formation

Much work has been done to expand the glass composition region available for operation of the Hanford Waste Treatment and Immobilization Plant. This includes the development of updated glass property-composition models as well as constraints. This report supports this effort by suggesting constraints for avoiding excessive, and likely detrimental, crystallization during melter operation while processing advanced low-activity glass waste forms. The constraints target SnO 2 and ZrO 2 crystals that can form when melter temperatures drop below 1100 °C. These types of crystals were found to be potentially detrimental during processing as they are denser than low-activity waste glass melts. SnO 2 , density of 6.95 g/cm3, and ZrO 2 , density of 5.68 g/cm3, have the potential to form during melter idling and settle to the bottom of the less dense glass melt (approximate density 2.65 g/cm3). If the crystals are present in appreciable amounts, they can result in blockages of the pour-spout riser, which impacts glass pouring and melter operation. Using previously acquired results and results from testing during this effort, constraints were determined and are suggested as options to reduce the risk of forming crystals of the types and concentrations that are likely detrimental to melter operation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Molten Salt Reactor Engineering Study for Off-Gas Management

In previous work published by the current team, the waste processing and waste form options were summarized for molten salt reactors (MSR) (Riley et al., 2018b; Riley et al., 2019). The primary types of waste from an MSR are summarized in Figure S1 and include (1) off-gas streams, (2) salt waste streams, (3) separated salt streams, (4) metal waste streams, (5) carbon waste streams, (6) decommissioning and decontaminating (D&D) waste streams, and (7) operating waste streams, or those that are generated from maintenance procedures. The primary focus of this report will be on management of the off-gas stream as it represents the pressure boundary for fuel-salt MSRs and, thus, is required for fission product confinement during reactor operation and reduction of the source term during a reactor accident.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2026

October 2025 Semiannual Composite Salt Waste Processing Facility Decontaminated Salt Solution Toxicity Characteristic Leaching Procedure Results

The aqueous waste from the Salt Waste Processing Facility is sampled semiannually for transfers to the Saltstone Disposal Facility. Salt solution is treated at the Saltstone Production Facility and disposed of in the Saltstone Disposal Facility. Per request of customer, X-TTR-Z-00027, Revision 0, one Saltstone Processing Facility Decontaminated Salt Solution and Z area premix material for the October 2025 semiannual Toxicity Characteristic Leaching Procedure sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request prepared by Savannah River Mission Completion. At 51 days cured, the Saltstone sample was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure. The October 2025 Semiannual Cement-Free grout sample met 61 79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act metals and Underlying Hazardous Constituents and met the Saltstone Production Facility Waste Acceptance Criteria that was in effect at the time of the sample collection at the Salt Waste Processing Facility.

Hsieh, Madison [Savannah River National Laboratory↗

DOE OSTI2021

Characterization of the Sulfur-Saturated Melt Versions of the HFG1 Study Glasses

The U.S. Department of Energy (DOE) is responsible for building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. The Office of River Protection (ORP) has requested that the Savannah River National Laboratory (SRNL) contribute to current glass studies to support successful startup of the WTP, due to SRNL’s recognized capabilities and expertise for glass waste form development. As part of this effort, the Pacific Northwest National Laboratory (PNNL) is developing, batching, and fabricating simulated high-level waste (HLW) glasses to generate the associated property data needed to validate or identify areas of potential characterization improvements in the current glass property/ composition models. Currently, the models cover only a small fraction of the waste compositions projected in the Hanford tank farm; therefore, the models must be expanded to include high fluoride compositions to successfully complete the WTP mission. SRNL support of this work is defined in the Task Technical and Quality Assurance Plan (TTQAP). This report provides results from the chemical analyses of a series of sulfur-saturated melt (SSM) versions of simulated nuclear waste glasses fabricated at PNNL. The glasses were selected as part of a broader study of the influence of glass composition on chemical durability, sulfur retention, and other properties. The glasses were designated the High Fluoride Glasses-1 (HLG1).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗