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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.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Glass-contact refractory of the nuclear waste vitrification melters in the United States: a review of corrosion data and melter life

The performance of the refractory lining in glass melters used for nuclear waste vitrification is critical to the melter reliability for long-term continuous operation. Monofrax® K-3, a high Cr 2 O 3 fused cast refractory material, has been widely used to build the liners of nuclear waste glass melters in the United States. Corrosion behaviour of Monofrax® K-3 refractory has been evaluated based on crucible-scale testing, inspection of the refractory components following scaled melter testing, and inspections of the Defense Waste Processing Facility (DWPF) melter refractory after service. The literature generally consists of empirical models based on short-term testing to describe refractory corrosion dependence on glass composition. Corrosion data from tests with longer testing times, at various temperatures, in the presence of molten salts, and with different redox reactions in the plenum atmosphere exist, may be insufficient to provide accurate refractory service life estimates. Additionally, the corrosion data collected under actual and scaled melter operating conditions are limited. Recommendations to achieve more direct correlation between the laboratory refractory corrosion data predictions and the observed melter service life are discussed to allow for more accurate predictions of the useful life of melter refractory linings.

Jin, Tongan↗
DOE OSTI2025

Dynamic and static corrosion of chromium-alumina refractory in simulated nuclear waste glass

Monofrax® K-3, a chromium–alumina refractory, is widely used in nuclear waste glass melters due to its high durability. However, aggressive vitrification conditions still lead to corrosion that limits melter lifespan. Here, this study investigates two key corrosion modes, subsurface and melt-line corrosion, under static and dynamic conditions, using a custom setup enabling precise control of melt velocity and temperature. Experimental results, interpreted using diffusion-limited dissolution models, show that (i) subsurface corrosion increases by approximately 10% per mm/s increase in melt velocity, while melt-line corrosion remains virtually unaffected, (ii) corrosion rate dependence on temperature is inversely proportional to melt viscosity, and (iii) glass composition affects both corrosion modes similarly, based on Cr 2 O 3 , Al 2 O 3 , and alkali contents. The experimental data also allowed the determination of diffusion coefficients of major dissolving species, which were found to be consistent between the melt-line and subsurface corrosion models, indicating their suitability for future CFD studies.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2021

Effect of water vapor and thermal history on nuclear waste feed conversion to glass

Water affects the glass melting process by interacting with the foam layer at the glass melt surface and by influencing the batch conversion reactions. Water vapor from the nuclear waste slurry feed maintains a high water vapor pressure in the melter atmosphere. To investigate to what extent water vapor affects the vitrification of nuclear waste in joule-heated, cold-top melters, a series of feed expansion experiments were performed under humid and dry atmospheres using samples of low-activity waste (LAW) melter feed simulants. Melting of feed pellets in the presence of water vapor slightly decreased the temperature of primary foam onset, but did not significantly affect the feed volume expansion by foaming or the foam-collapse temperature. Sets of feed expansion experiments and evolved gas analyses were also performed to check the effect of thermal history on LAW feed samples tested as direct slurry, loose powder, and slow- and fast-dried pellets.

Marcial, Jose↗
DOE OSTI2020

Cask RT-100 Type B(U) Self-Certification for DOT 7A, Type A Nuclear Waste Transportation - 20098

The purpose of this Type A certification is to diversify the use of the RT-100 cask by allowing the end user to transport either Type A or Type B radioactive waste in accordance with the handling procedures for each waste category. The project's scope included the development of a matrix that contains all applicable requirements for a Type A package from the Code of Federal Regulations (CFR), Title 49 for the Department of Transportation (DOT) with a justification for each requirement. Moreover, the scope included the writing of conditions and restrictions list for the transportation of Type A nuclear waste, the writing of a compliance statement, and the design and fabrication of a Type A placard to be permanently placed on the RT-100 cask. Type A packages are self-certified by the end user using one of three methods as permitted by 49 CFR 173.461, either by performance of tests with prototypes or samples of the specimens representing the packaging, or reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature, or calculations or reasoned evaluation using reliable and conservative procedures and parameters. Since the RT-100 cask was licensed by the Nuclear Regulatory Commission (NRC) as a Type B cask, Robatel relied on the method of reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature for Type A certification. This method was determined to be acceptable, since Type B packages are evaluated for more demanding transport conditions than those required for Type A packages. This paper presents Robatel's process for certifying the RT-100 cask for Type A shipments, an overview of the DOT regulations related to Type A packages, and the documents that were generated for the end user. This paper is concluded with a discussion on the advantages of having the option to transport both Type A and Type B nuclear waste, and possible future projects to further improve the operation of the cask when making Type A shipments. (authors)

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

Efficient Multi-Scale, Thermo-Hydro-Mechanical Numerical Model for Simulation of Long-Term Stability of Rooms for Nuclear Waste Disposal

This project demonstrated the feasibility of developing a numerical modeling tool that provides an efficient and unique approach to analyze the geomechanical performance of a deep geological repository (DGR) for high-level nuclear waste. The numerical modeling tool is physics-based and models the thermo-hydro-mechanical (THM) processes during the required time (up to one million years) for a DGR. The unique capability of the developed code is explicit representation of fracturing and a discrete fracture network (DFN) in the model that includes the entire repository layout for simulated time of one million years. Therefore, it can potentially address one of the challenges of analysis of a DGR, which is efficient numerical simulation of a multi-scale, physics-based models.

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

Nanohybrid of Silver‐MXene: A Promising Sorbent for Iodine Gas Capture from Nuclear Waste

The increasing reliance on nuclear energy as a significant low-carbon power source necessitates effective solutions for managing radioactive emissions. This study introduces a novel application of MXene nanohybrids, specifically silver-MXene (Ag-Ti 3 C 2 T x ), as an effective sorbent for radioiodine off-gas capture at an operating temperature of 150 °C. Through comprehensive material characterization, including X-ray diffraction, scanning and transmission electron microscopies, energy-dispersive X-ray spectroscopy, Raman spectroscopy, thermogravimetric analysis, inductively coupled plasma optical emission spectroscopy, and gas sorption analyses, the successful loading of Ag nanoparticles onto Ti 3 C 2 T x is confirmed and the subsequent formation of AgI upon iodine capture. The results demonstrate that Ag-Ti 3 C 2 T x exhibits superior iodine uptake compared to traditional silver-based sorbents such as silver mordenite zeolite (AgZ) and silver-functionalized silica aerogel (AgAero). The Ag-Ti 3 C 2 T x achieves an iodine loading of 946 mg g −1 , significantly outperforming AgZ (131 mg g −1 ). These findings highlight the potential of Ag-Ti 3 C 2 T x as a highly efficient, thermally stable sorbent for radioiodine capture, and potentially addressing key limitations of existing materials.

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

Uncertainty propagation and sensitivity analysis for constrained optimization of nuclear waste vitrification

Abstract The vitrification of high‐level waste (HLW) by heating a mixture of glass‐forming chemicals (GFCs) with the waste can be improved using a constrained optimization problem. This study explores how different uncertainty propagation (UP) methods implemented with the optimization process can affect the glass formulation of nuclear waste glasses. UP is the effort of propagating uncertain inputs through a system to understand and quantify output distributions. Uncertainty intervals are crafted from output distributions to inform the optimization algorithm. UP is often implemented with Monte Carlo (MC) sampling for large nonlinear systems, which can be difficult to implement within a constrained optimization algorithm that requires derivative information. Other UP methods often used for optimization under uncertainty (OUU) can be designed to work within an established constrained optimization framework. Methods of UP are evaluated in this study including iterative sampling approaches, first‐order approximations, and surrogate modeling with machine learning (ML). A method of dimensional reduction based on global sensitivity analysis is introduced to support the UP methods for the large dimensionality of the problem. Analytical UP methods able to achieve similar optimums 10 times faster than the baseline MC approach, and produce 93.9% similar output distributions are reported.

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

Thermophysical properties of the Ghareb formation relevant for nuclear waste disposal

The Ghareb formation, a shallowly buried porous chalk in Israel, is currently a candidate for nuclear waste disposal. The potential repository is somewhat unique for its host rock and emplacement in shallow (500 m) engineered large diameter boreholes. Herein, the thermal properties of the Ghareb are determined to support design and performance assessment; the relevant properties measured are thermal conductivity, specific heat capacity, thermal diffusivity, and thermal expansion coefficient along with their relationship with varying temperature. For the temperature range of 40 to 275 °C, the thermal conductivity ranges from 0.30 to 1.10 W/m·K, the thermal diffusivity ranges from 0.20 to 0.72 mm 2 /s, and volumetric heat capacity ranges from 0.86 to 2.00 MJ/m 3 ·K. Thermal strain measurements were used to estimate the linear thermal expansion coefficient to be 6·10 –4 –9·10 –2 °C -1 from 40 to 300 °C. These measured properties were used in a thermomechanical model to estimate near-field stresses an hour and 10 years after waste emplacement; the borehole was found to be stable. Thermal loading after 10 years was predicted to elevate local pore pressures by 1–1.5 MPa. The laboratory measurements coupled with analyses are the first attempts at performance assessment characterization for this first of its kind potential repository setting with this chosen host rock.

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

Modeling nuclear waste disposal in crystalline rocks at the Forsmark and Olkiluoto repository sites – Evaluation of potential thermal–mechanical damage to repository excavations

We conduct coupled thermo-hydro-mechanical modeling of a KBS-3V repository design in crystalline rocks, using data and conditions from the Forsmark in Olkiluoto repository sites in Sweden and Finland. The study focuses on repository performance related to the impact of thermal and hydraulic evolution on the potential for thermal–mechanical damage to underground repository excavations. For the designs and conditions considered at the Forsmark and Olkiluoto repository sites, the simulations show a peak temperature well under the adopted performance target of a 100°C maximum temperature, whereas there is still a high potential for thermal–mechanical damage to the KBS-3V waste deposition holes. The thermal–mechanical damage is much more likely if rock permeability is so low that it delays saturation and swelling of bentonite-clay-based backfill beyond the time for the thermal–mechanical peak, which occurs 50 to 100 years after nuclear waste deposition. We also found that sidewalls of the KBS-3V emplacement tunnels are vulnerable to tensile fracturing due to the combined effect of thermal stressing and backfill swelling. The study highlights a strong interaction between bentonite-based backfill and host rock through capillary suction along with induced rock desaturation. A careful design and selection of the bentonite-clay-based backfill materials for KBS-3V tunnels and deposition holes can facilitate a timely saturation and backfill swelling that in turn can minimize thermal–mechanical damage.

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

Real-time infrared spectroscopy coupled with blind source separation for nuclear waste process monitoring

On-line infrared absorbance spectroscopy enables rapid measurement of solution-phase molecular species. Many spectra-to-concentration models exist for spectral data, with some models able to handle overlapping spectral bands and nonlinearities. However, model accuracy is limited by the quality of training data used in model fitting. The process spectra of nuclear waste simulants at the Savannah River Site display incongruity between training and process spectra; the glycolate spectral signature in the training data does not match the glycolate signature in Savannah River National Laboratory process data. A novel blind source separation algorithm is proposed that preprocesses spectral data so that process spectra more closely resemble training spectra, thereby improving model quantification accuracy when unexpected sources of variation appear in process spectra. The novel blind source separation preprocessing algorithm is shown to improve nitrate quantification from an R 2 of 0.934 to 0.988 and from 0.267 to 0.978 in two instances analyzing nuclear waste simulants from the Slurry Receipt Adjustment Tank and Slurry Mix Evaporator cycle at the Savannah River Site.

Crouse, Steven H.↗
DOE OSTI2020

The dissolution of simulant vitrified intermediate level nuclear waste in young cement water

Abstract It is pertinent to the safety case for geological disposal in the UK that the behaviour of vitrified wastes in proximity to cementitious materials is understood. In this study, vitrified simulant intermediate level nuclear waste (ILW) was subject to dissolution in a synthetic cement water solution to simulate disposal conditions. Results show that the presence of alkali / alkaline earth elements in the cementitious solution can be favourable, at least in the short-term, leading to lower dissolution rates associated with incorporation of these elements into the altered layer of the glass.

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

Energetics of Salt-Bearing Sodalites, Na 8 Al 6 Si 6 O 24 X 2 (X = SO 4 , ReO 4 , Cl, I): A Treatment Option for Pertechnetate-Enriched Nuclear Waste Streams

An alternative option for treating anion-enriched reprocessed nuclear waste streams is to immobilize technetium-99 ( 99 Tc, β = 293.7 keV, t 1/2 = 2.1 × 10 5 years) and other anions in micro- and mesoporous materials. Here we determine the thermodynamic stability of anion bearing sodalites, Na 8 Al 6 Si 6 O 24 X 2 (X = SO 4 , ReO 4 , Cl, I), to improve our understanding of the driving forces that control framework assembly using high temperature oxide melt solution calorimetry. Raman and FTIR spectroscopy illustrate a strong dependence for vibrational features on anion size and enabled the development of a linear model that predicted the vibrational features for numerous anion bearing sodalites to within ±20 cm –1 (i.e., OH, F, Br, ClO 4 , NO 3 , and MnO 4 ). The largest negative enthalpy of formation from elements and the lack of structural water demonstrate that the perrhenate sodalite (Na 8 Al 6 Si 6 O 24 [ReO 4 ] 2 ), a chemical analogue for pertechnetate sodalite (Na 8 Al 6 Si 6 O 24 [TcO 4 ] 2 ), is more thermodynamically stable than all other anion bearing sodalites evaluated. The enthalpies of the reaction between nepheline and the sodium salt, which provides the guest anion species, was negative only for the ReO 4 and NO 3 bearing sodalites. Overall, we report for the first time the enthalpy of the ion exchange reactions for different anion bearing sodalites relative to the perrhenate sodalite, which is a key step in gaining the ability to tune sodalite material properties and structure during treatment and the immobilization of 99 Tc in the presence of competing anions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Structure–property relations of sodium iron phosphate nuclear waste glasses: Effects of iron redox ratio and glass composition

Iron phosphate glasses, known for their exceptional chemical durability and potential applicability in nuclear waste management, have gained significant attention over the years. The structures of these glasses are complicated by the coexistence of Fe 3+ and Fe 2+ , which plays a crucial role in determining their structures and properties. Here, this work uses molecular dynamics simulations to study the structural changes in Na 2 O–Fe 2 O 3 –P 2 O 5 glasses with varying glass composition and Fe 2+ /Fe 3+ redox ratio. It was found that the redox ratio and modifier contents significantly affected the short-range and medium-range orders in the glasses. Significant changes in the local environments around P 5+ and Fe 3+ were observed, as reflected by the bond distances and coordination numbers. Na + cations are found to preferentially associate with Fe 3+ (rather than Fe 2+ ), whereas Fe 2+ has stronger association with P 5+ than Na + , confirming the structural role of Fe 2+ as a glass modifier. The disruptions in P–O–P linkages upon increasing FeO suggest that FeO causes glass depolymerization. These glasses achieved higher connectivity with increasing Fe 3+ / (Fe 3+ + Fe 2+ ) ratios, conerting phosphorous Q 2 to Q 3 units and iron Q 5 units to Q 4 units. The decrease of nonbridging oxygen fractions with increasing Fe 3+ / (Fe 3+ + Fe 2+ ) ratios, through creating P–O–Fe linkages, is the main reason of enhanced network connectivity. Quantitative structure–property relationship analyses with different structural descriptors were used to correlate with measured properties. The analyses provided valuable insights into structure–property relationships, emphasizing the importance of choosing relevant energy parameters and defining glass network connectivity, particularly in F net descriptors. It was found the Fe–O–P linkage density exhibits strong correlations to measured dissolution rates, supporting the importance of these linkages in improving the chemical durability in iron phosphate glasses.

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

Unveiling the effect of composition on nuclear waste immobilization glasses’ durability by nonparametric machine learning

Abstract Ensuring the long-term chemical durability of glasses is critical for nuclear waste immobilization operations. Durable glasses usually undergo qualification for disposal based on their response to standardized tests such as the product consistency test or the vapor hydration test (VHT). The VHT uses elevated temperature and water vapor to accelerate glass alteration and the formation of secondary phases. Understanding the relationship between glass composition and VHT response is of fundamental and practical interest. However, this relationship is complex, non-linear, and sometimes fairly variable, posing challenges in identifying the distinct effect of individual oxides on VHT response. Here, we leverage a dataset comprising 654 Hanford low-activity waste (LAW) glasses across a wide compositional envelope and employ various machine learning techniques to explore this relationship. We find that Gaussian process regression (GPR), a nonparametric regression method, yields the highest predictive accuracy. By utilizing the trained model, we discern the influence of each oxide on the glasses’ VHT response. Moreover, we discuss the trade-off between underfitting and overfitting for extrapolating the material performance in the context of sparse and heterogeneous datasets.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Reduced-order modeling of near-field THMC coupled processes for nuclear waste repositories in shale

Performance assessment (PA) of geologic radioactive waste repositories requires three-dimensional simulation of highly nonlinear, thermo-hydro-mechanical-chemical (THMC), multiphase flow and transport processes across many kilometers and over tens to hundreds of thousands of years. Integrating the effects of a near-field geomechanical process (i.e. buffer swelling) into coupled THC simulations through reduced-order modeling, rather than through fully coupled geomechanics, can reduce the dimensionality of the problem and improve computational efficiency. In this study, PFLOTRAN simulations model a single waste package in a shale host rock repository, where re-saturation of a bentonite buffer causes the buffer to swell and exert stress on a highly fractured disturbed rock zone (DRZ). Three types of stress-dependent permeability functions (exponential, modified cubic, and Two-part Hooke’s law models) are implemented to describe mechanical characteristics of the system. Overall, our modeling study suggests that compressing fractures reduces DRZ permeability, which could influence the rate of radionuclide transport and exchange with corrosive species in host rock groundwater that could accelerate waste package degradation. Less permeable shale host rock delays buffer swelling, consequently retarding DRZ permeability reduction as well as chemical transport within the barrier system.

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

Nanoscopic structure of borosilicate glass with additives for nuclear waste vitrification

In this report we investigated the nanoscopic structure of borosilicate glasses as a host for high-level radioactive liquid waste (HLLW) in the presence of the additives Na 2 O and CaO/ZnO with and without Li 2 O. These additives have been used to lower the glass melting point, suppress the macroscopic phase separation, and increase the chemical durability of glasses. Small-angle neutron scattering was used to elucidate the effect of the additives on the nanoscopic structure along with macro- and atomic-scale observations obtained using neutron imaging and X-ray diffraction, respectively. The borosilicate glass made without Na 2 O formed a microphase-separated structure of SiO 2 - and B 2 O 3 -rich domains approximately 100 nm in size, and the material was homogeneous at length scales up to millimeters. With the addition of increasing amounts of Na 2 O, the domain spacing decreased. Introducing CaO/ZnO additives induced inhomogeneities in the glass, such as void structures several nanometers in size, although the inhomogeneity was suppressed by the coaddition of Li 2 O. These results provide insights into the performance of host glasses for HLLW vitrification. Incorporating HLLW into glasses is likely to cause changes in the nanoscopic structure of host glasses similar to those revealed here.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗