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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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Identification and time evolution of thionyl chloride (SOCl 2 ) radiolysis products

In this study, innovative solutions are needed to reduce the amount of high-level waste generated by used nuclear fuel recycling strategies to support the widespread adoption of sustainable nuclear fission energy technologies. To this end, a new sulfur chloride-based process has been developed to recycle zirconium alloy-based materials, which make up a significant fraction of high-level radioactive waste. To support the continued development of this process, we present new data on the potential reaction pathways over time of the products arising from the gamma and electron beam radiolysis of neat thionyl chloride (SOCl 2 ). Interrogation of the gamma irradiated liquid by Raman spectroscopy provided more conclusive identification of the SOCl 2 degradation products, specifically sulfur dichloride (SCl 2 ), molecular chlorine (Cl 2 ), sulfur dioxide (SO 2 ), and sulfuryl chloride (SO 2 Cl 2 ). In comparison, the high dose rate (~10 7 Gy s -1 ) electron beam irradiations formed significantly more degradation products. For both cobalt-60 gamma and electron beam irradiations, the observed degradation products were found to evolve as a function of time post-irradiation via the same reaction pathways, with indication of a solvent regeneration mechanism. These findings are fortuitous for process development, as such a mechanism would be beneficial for process longevity and cost effectiveness.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Advances in Resin Management Using 3R-Scan - 20154

The most important factor underlying optimal waste management is developing a clear picture of the radioactivity content of the waste and its impact on waste disposal cost. For over 35 years since the publishing of 10CFR61, waste characterization has relied on sampling the final waste product after formation. In the days following 10CFR61, the cost of final disposal was marginal with only a small impact on the overall costs. Constraints were added by provisions of the Low Level Waste Policy Act of 1985 leading to increasingly limited access to those disposal sites that remained available. In addition, Nuclear Regulatory Commission (NRC) pressure promoting waste volume reduction led to disposal costs inevitably rising. Despite this, characterization practices in monitoring of waste generation for activity content still center on the same dated processes. This results in a disposal classification on the basis of endpoint sampling without consideration of the homogeneity of the waste mixture. It can also disregard consideration of the representativeness of the single or small sample base. As a minimum effort, a formalized sampling program of a fixed grouping of waste streams can be implemented that could account for more than 95% of all of the activity carried in solid waste products. The sample results for each radionuclide could then be trended as time passes to develop reasonable scaling factors for difficult to measure radionuclides. This process, identified in NRC guidance, has been rigorously followed by a relatively small number of facilities. The trended scaling factors serve to improve accuracy by identifying anomalous results that could otherwise go undetected. Direct monitoring of the accumulation of activity in process streams generating solid radwaste, including demineralizers and filter streams, is a more precise approach. Nearly all of these streams are monitored by plant chemistry on a regular schedule to maintain water quality. This paper discusses viable options for developing the basis for characterization through process monitoring of the accumulation of activity at the point of generation. Special focus is on resin bed tracking and how process knowledge of these streams can be brought together to form a consistent and precise solid waste radioactivity inventory. Some of the specific points covered in this paper include the merger of the fission product release computer program, 3R-STAT, with the radwaste sample analysis computer program, SCAN4 to create 3R-SCAN, the importance of individual waste stream influences on the overall source term, and the use of historic sample data to develop scaling factors using an automated process. (authors)

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

Implementation and verification of a user-defined element (UEL) for coupled thermal-hydraulic-mechanical-chemical (THMC) processes in saturated geological media

Efficient and accurate modeling of the coupled thermal-hydraulic-mechanical-chemical (THMC) processes in various rock formations is indispensable for designing energy geo-structures such as underground repositories for high-level nuclear wastes. Here, this work focuses on developing and verifying an implicit finite element solver for generic coupled THMC problems in geological settings. Starting from the mass, momentum, and energy balance laws, a specialized set of governing equations and a thermoporoelastic constitutive model is derived. This system is then solved by an implicit finite element (FE) scheme. Specifically, the residuals and the Jacobians are scripted in a user-defined element (UEL) subroutine which is then combined with the general-purpose FE software Abaqus Standard to solve initial-boundary value problems. Considering the complexity of the system, the UEL development follows a stepwise manner by first solving the coupled hydraulic-mechanical (HM) and thermal-hydraulic-mechanical (THM) equations before moving on to the full THMC problem. Each implementation step consists of at least one verification test by comparing computed results with closed-form analytical solutions to ensure that the various coupling effects are correctly realized. To demonstrate the robustness of the algorithm and to validate the UEL, a three-dimensional case study is performed with reference to the in-situ heating test of ATLAS at Belgium in 1980s. A hypothetical radionuclide leakage event is then simulated by activating the chemical-concentration degree of freedom and prescribing a constant high concentration at the heater's surface. The model predicts a limited contaminated regime after six years considering both diffusion and advection effects on species transport.

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

Performance Assessment for the E-Area Low-Level Radioactive Waste Disposal Facility at the Savannah River Site: Appendix H

This section provides supporting material for the development of DU-specific final inventory limits for the GW pathways for every generic waste form and SWF parent radionuclide requiring an inventory limit. The final inventory limits are based on nominal PA transport simulations using PORFLOW as reported in Chapter 5. The nominal PA settings represent the compliance case where some modeling parameter settings are defined based on conservative (pessimistically leaning) arguments. In the overall computational approach employed in this PA, a multitiered radionuclide screening process is adopted as discussed in Chapter 2, Section 2.3.7. In the multitiered process, the initial list of 1,252 parent radionuclides is shortened substantially using conservative, but simple, transport models, along with a reasonably low cutoff criterion of 0.1% SOF value. Multidimensional PORFLOW flow and transport modeling is employed for every parent radionuclide that failed the GW screening. The generic waste form limits represent Tier-3 analyses, while Tier-4 analyses are employed for SWF limits, where warranted.

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

Monsanto Research Corp Model 2493 Series Special Form Re-Evaluation

The purpose of this memo is to characterize Monsanto Research Corp. (MRC) model 2493 series radioactive sealed sources as US Department of Transportation (DOT) Special For Radioactive material. with the goal of achieving shipment, consolidation, interim storage, and final disposition.

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

Measurement of Iodine, Iodide, and Iodate in Hanford Tank Waste: Technology Transfer from PNNL to Hanford 222-S Laboratory

This report summarizes the measurement of the inorganic chemical forms of iodine in Hanford tank waste. The Hanford tank wastes have roughly a part per million iodine. Part of the iodine is the long-lived beta emitter 129I, and the rest is stable 127I. Because tank waste iodine is radioactive and radiotoxic, its chemical form must be known so that its behavior in the vitrification process can be reliably known and anticipated in various waste streams (glass, secondary liquid and solid waste, and offgas emissions) at Hanford. Iodine in tank wastes exists in the inorganic chemical forms iodide and iodate (and possibly also periodate) and also as organic forms such as alkyl iodides. The measurement of organic iodides is very different from the measurement of inorganic forms and is outside the scope of this report. The inorganic chemical forms are chemically separated, in sequence, from the raw tank waste using solvent extractions and several redox reactions, then measured by ICP-MS. The inorganic chemical forms of iodine are chemically reactive, and so is the tank waste. The chemistry must be carefully designed to avoid unintended reactions that could convert one form of iodine to another during the analysis, which would skew the data and report iodine in the wrong chemical form. This iodine analysis, developed at PNNL, is being transferred to the DOE 222-S Laboratory on the Hanford Site, in Washington State, for implementation during waste processing operations.

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

Plutonium Oxidation State Distribution in the Presence of WIPP-Relevant Organics and Iron Corrosion Products

The oxidation state and solubility of plutonium (Pu) in high ionic strength synthetic WIPP (Waste Isolation Pilot Plant) brines as a function of pC H+ in the presence and absence of WIPP-relevant organic ligands (EDTA [Ethylenediaminetetraacetic acid], oxalate, citrate, acetate) and iron corrosion products (magnetite and metallic iron) at 𝑇 = 23 ± 2 ∘C was thoroughly studied by long-term batch solubility experiments (between approximately 800-1,100 days) from an undersaturation approach. The oxidation state of Pu in the WIPP environment has been a topic of interest since the initial Compliance Certification Application (CCA). This study aims to investigate the solubility of Pu under the expected WIPP conditions and to determine the oxidation state of the solid phase that will control the solubility. One of the most important results of this study is that the Pu oxidation state was analyzed both from the surface area of the corrosion products and in the precipitated solid. The analysis of the Pu oxidation state on the surface of the iron mineral from the ongoing undersaturated experiments is more relevant to the performance assessment of nuclear waste disposal than short term batch (plutonium-iron phase) experiments. The X-ray Absorption Near-Edge Spectroscopy (XANES) analysis showed that Pu oxidation state is different on the metal surface and in the precipitated solid. Pu(III) is the dominant oxidation state in the metallic iron (Fe 0 ) system in the presence and absence of organics. Pu(IV) is the dominant oxidation state in the magnetite system in the presence and absence of organics. Also, organics stabilize Pu(IV) in the magnetite system. Analysis of Pu in the precipitated solid by Extended X-ray Absorption Fine Structure (EXAFS) analysis showed that Pu formed an inner-sphere complex with iron with minor amounts of PuO 2 present. X-ray diffraction (XRD) results indicate that metallic iron and magnetite did not oxidize in three years in the alkaline and high ionic strength system. Under these conditions (8 < pC H+ < 10 at T = (22 ± 2) °C under nitrogen atmosphere, the solubility of Pu changes by up to three orders of magnitude (10 -5 and 10 -8 M). The spread in solubility is highest at pC H+ = 9. Pu(III) and Pu(IV) showed different solubility behavior in the synthetic WIPP brine. A summary of the data collected in this report will be submitted to Sandia National Laboratories as part of a parameter update report which will outline the changes to the OXSTAT parameter for the 2026 Compliance Recertification Application (CRA-2026). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Effects of Radiolysis, Organic Complexation, and Redox Conditions on the Speciation and Oxidation State Distribution of Pu(III/IV)” (LCO-ACP-25). All data reported were obtained under the Los Alamos National Laboratory-Carlsbad Office (LANL-CO) Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD).

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

Spent Nuclear Fuel Receipt Rate Analysis within an Integrated Waste Manage-ment System Architecture that Includes Consolidated Interim Storage

A key parameter in analyzing the performance of an integrated waste management system (IWMS) architecture for the disposition of spent nuclear fuel (SNF) is the SNF receipt rate from reactor and other custodian sites. The introduction of one or more federal consolidated interim storage facilities (CISFs) into the IWMS architecture can enable the receipt rate profile as a function of time to be accelerated relative to system architectures without a CISF. The question then arises as to what an optimal SNF receipt rate profile for an IWMS architecture might be in view of practical constraints and desired system performance attributes and associated metrics. This paper describes a sensitivity study on SNF receipt rates and the associated results for a selected set of IWMS scenarios aimed an in-forming near-term planning for interim storage capabilities and transportation assets. Two different strategies are compared, one that fills an initial CISF quickly and then idles the transportation system while a disposal system is prepared, and a second strategy that aims to provide a more continuous use of transportation assets and receipt capabilities at the IWMS while the disposal system is readied for SNF receipt. Cost considerations and other factors such as impact on timing of clearing reactor sites of SNF, efficient use of capital assets, and other metrics, including those which may be important to a CISF host community, are examined. Based on the analysis, an initial approach is presented targeting a continuous receipt strategy while having the flexibility to step up receipt capabilities to a reasonable degree when needed and beneficial within overall system constraints.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2021

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY21 Status Report)

This progress report describes work performed during FY21 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of canister materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In FY21, modeling and experimental work was performed that further defined our understanding of the potential chemical and physical environment present on canister surfaces at both marine and inland sites. Research also evaluated the relationship between the environment and the rate, extent, and morphology of corrosion, as well as the corrosion processes that occur. Finally, crack growth rate testing under relevant environmental conditions was initiated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2002

Fusion for Space Propulsion

The need for fusion propulsion for interplanetary flights is discussed. For a propulsion system, there are three important system attributes: (1) The absolute amount of energy available, (2) the propellant exhaust velocity, and (3) the jet power per unit mass of the propulsion system (specific power). For efficient and affordable human exploration of the solar system, propellant exhaust velocity in excess of 100 km/s and specific power in excess of 10 kW/kg are required. Chemical combustion obviously cannot meet the requirement in propellant exhaust velocity. Nuclear fission processes typically result in producing energy in the form of heat that needs to be manipulated at temperatures limited by materials to about 2,800 K. Using the fission energy to heat a low atomic weight propellant produces propellant velocity of the order of 10 kinds. Alternatively the fission energy can be converted into electricity that is used to accelerate particles to high exhaust velocity. However, the necessary power conversion and conditioning equipment greatly increases the mass of the propulsion system. Fundamental considerations in waste heat rejection and power conditioning in a fission electric propulsion system place a limit on its jet specific power to the order of about 0.2 kW/kg. If fusion can be developed for propulsion, it appears to have the best of all worlds - it can provide the largest absolute amount of energy, the propellant exhaust velocity (> 100 km/s), and the high specific jet power (> 10 kW/kg). An intermediate step towards fusion propulsion might be a bimodal system in which a fission reactor is used to provide some of the energy to drive a fusion propulsion unit. There are similarities as well as differences between applying fusion to propulsion and to terrestrial electrical power generation. The similarities are the underlying plasma and fusion physics, the enabling component technologies, the computational and the diagnostics capabilities. These physics and engineering capabilities have been demonstrated for a fusion reactor gain (Q) of the order of unity (TFTR: 0.25, JET: 0.65, JT-60: Q(sub eq) approx. 1.25). These technological advances made it compelling for considering fusion for propulsion.

Thio, Y. C. Francis↗
DOE OSTI2020

Lessons Learned in the Development of Industry Guidance - 20279

The availability of industry guidance on different aspects of waste management forms a valuable reference for waste practitioners. The guidance typically details the principles, processes and practices that should be consulted during each stage of the relevant activities. It also includes relevant regulatory and international requirements, appropriate case studies and examples of best practices to illustrate technical concepts and make the information available to the reader in an accessible manner. Over recent years Wood have developed (or are developing) a number of industry guidance documents on a range of topics including Solid Waste Characterization, Routine Water Quality Monitoring on Nuclear Sites: Pipeline Characterization: Determining Background Radioactivity for Land Quality Assessment: Interim Storage and Aqueous Waste Management. This talk provides an overview of lessons learned along the way including: - the importance of stakeholder engagement through the process: - how to identify 'good practice'; - the importance of iterative development; and - the selection of illustrative and informative case studies. (authors)

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

Structural Changes in Molten Salt Fuel and/or Waste Stream Compounds Cs 2 UCl 6 and Cs 2 UO 2 Cl 4 from Room Temperature to Melting and Related Materials [Slides]

Perovskite phases have been explored in recent literature as waste forms for long term storage of radioactive waste from molten salt reactors specifically. Synthesis of new halide perovskite phases can inform these efforts as well as molten salt reactor chemistry in general. Inorganic perovskites have also shown promise as new scintillating materials for detection of X-rays and gamma rays. Modifying the structure and chemistry of these compounds, perhaps with organic components, could be used to synthesize new scintillating compounds or modify the properties of existing ones. Both U 6+ and U 4+ readily form compounds with a striking similarity to vacancy ordered perovskites (Cs 2 UO 2 Cl 4 and Cs 2 UCl 6 ). Each of these compounds is relevant to molten salt reactor chemistry and/or waste streams. These phases could from in uranium chloride salt systems or waste streams: Cs being a fission product.

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

Transformation of Ferrihydrite to Goethite and the Fate of Plutonium

Understanding the interactions between plutonium and iron (oxy)hydroxide minerals is necessary to gain a predictive understanding of plutonium environmental mobility and to evaluate long-term performance of nuclear waste repositories. We investigated the fate of plutonium during the formation of ferrihydrite and its subsequent transformation into goethite. Ferrihydrite was synthesized with varying quantities of Pu(IV) following either a sorption or coprecipitation process; the ferrihydrite was then aged hydrothermally to yield goethite. Additionally, the synthesized materials were characterized via extended X-ray absorption fine structure spectroscopy, transmission electron microscopy, and acid leaching to elucidate the nature of plutonium association with ferrihydrite and goethite. In samples prepared following the sorption method, plutonium was identified in two different forms: a PuO 2 precipitate and a surface-sorbed plutonium complex. For the samples prepared via coprecipitation, no PuO 2 formation occurs in the ferrihydrite precursor and in the goethite experiments where plutonium concentration is ≤1000 ppm (mg kg –1 ). In these coprecipitation products, plutonium is strongly bound to the minerals either via formation of an inner sphere complex, or via an incorporation process. In the coprecipitation experiments, PuO 2 formation only occurs at the highest plutonium concentration (3000 ppm), suggesting that during ferrihydrite transformation into goethite, part of the plutonium can be remobilized to form PuO 2 nanoparticles. Collectively, our results demonstrate that the nature of plutonium associated with the precursor ferrihydrite (adsorbed vs coprecipitated) will have a direct impact on the association of plutonium with its alteration product (goethite). Furthermore, the data illustrate that some properties of plutonium association with the precursor ferrihydrite are retained through the transformation into goethite. These findings show that plutonium strongly associates with iron (oxy)hydroxides formed through coprecipitation processes and in these materials, plutonium can be strongly retained by the iron minerals.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Applying 3D Geologic Modeling Workflows to the Argillite Reference Case (Rev. 1)

The objective of this short report is to document the application of our 3D geologic modeling workflow to an argillite (shale) host rock. Over the past four years, our team at Los Alamos National Laboratory has developed a geologic modeling workflow that can be applied to generic alluvial basins such as those found in the western United States. In “frontier” or “exploratory” basins where data are sparse, the first steps are to collect, evaluate and integrate available subsurface data into conceptual geologic models. Those models form the basis for constructing the geologic framework model, a 3D geocellular model ideally constrained by seismic and borehole data. To date we have constructed our models using “synthetic” well data derived from conceptual models, without the prospect of validating our workflow using “real” subsurface data. We were tasked to investigate whether our workflow designed for alluvial basin sediments could be applied to other potential repository host rocks. This task also provided the opportunity to work with high-quality subsurface data collected specifically for siting and evaluating a nuclear waste repository. Nagra, the Swiss governmental agency responsible for the disposal of the nation’s radioactive waste, generously provided us with data from two deep boreholes drilled through their argillaceous target formation. The aim of our proof-of-concept demonstration is to evaluate whether geostatistical methods offer a viable approach to property modeling in argillaceous rocks. Nagra provided us with the well data on the condition that we maintain confidentiality with all transferred information and results. Fortunately, Nagra posts numerous technical reports on its public website that describe the subsurface geology in great detail. All of the information and illustrations in this report related to the Swiss repository enterprise are taken from the Nagra public website.

58 GEOSCIENCES↗
DOE OSTI2023

Increasing the Fissile Mass Loading of High-Level Waste Glass Canisters to Greater Than 2,500 g/m 3 at the Savannah River Site

To eliminate future fissile mass loading constraints for the Savannah River Site H-Canyon Facility and Liquid Waste system, the Savannah River National Laboratory recommends a repository evaluation of a uranium fissile mass loading in glass at 7,144 g/m 3 in addition to the existing International Atomic Energy Agency safeguards and security limit of 2,500 g/m 3 total plutonium. This recommended increase above the authorized 2,500 g/m 3 fissile mass loading limit is based on concentrations of uranium and plutonium shown to produce an acceptable glass waste form rather than projections of the maximum fissile mass loading in future sludge batches. An authorized fissile mass loading limit greater than 2,500 g/m 3 will increase facility flexibility, reduce the number of high-level waste canisters produced, and potentially avoid an increase in the Department of Energy Environmental Management mission life without compromising safety or glass product quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗