Search NASA⌕ Search

SEARCH · Search NASA

Results for “waste characterization and management”

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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Composition Measurements of the LAW ML1 Glasses

This report provides the results from the chemical analyses of the glass compositions of the quenched Low Activity Waste Machine Learning study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with target compositions. The relative differences between the target and measured concentrations of F - for one glass, P 2 O 5 for two glasses, SO 3 for three glasses, and ZrO 2 for two of the glasses were greater than 10%. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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

Glass Formulation and Testing with RPP-WTP HLW Simulants (Final Report)

This report presents the results from HLW glass formulation work that was conducted at the Vitreous State Laboratory of The Catholic University of America during Part B1 of the RPP-WTP Project (formally known as TWRS-P) in support of the Privatization contractor, BNFL, Inc. Glass formulation development is required to provide data that are essential for design of the vitrification facility in the near term and for safe and reliable operation of that facility in the longer-term. These data also form the bases from which contractually and technically acceptable immobilized high-level waste (IHLW) products will be devised. Glass formulation therefore impacts many aspects of the vitrification facility design, overall process economics, and contractual compliance. As an integral part of the overall system design, development of glass formulations requires input from many sources including tank waste characterization, pretreatment process definition, flow sheet development, and processing schedules. Accordingly, the glass formulation effort must be responsive to new information as the facility design basis evolves.

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

Composition Measurements of the LAW ALG Glasses

This report provides the results from the chemical analyses of glass compositions for the Low-Activity Waste Algorithm study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with targeted compositions. The relative differences between the targeted and measured concentrations of SO 3 for most of the glasses were greater than 10%. The relative differences between the targeted and measured concentration of F- was greater than 10% for one glass. The relative differences between the targeted and measured concentration of SnO 2 was greater than 10% for one glass. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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

Composition Measurements of the LAW Phase 5 Glasses

This report provides the results from the chemical analyses of glass compositions for the Low-Activity Waste Phase 5 study glasses, a series of simulated nuclear waste glasses fabricated at Pacific Northwest National Laboratory. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with targeted compositions. The relative differences between the targeted and measured concentrations of CaO, Cl - , K 2 O, Li 2 O, and SO 3 for some of the glasses were greater than 10%. Several of these glass samples were rerun to verify original measurements; however, there were no significant changes in the measurements to indicate errors in preparation or analysis of the samples. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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

Status of the Characterization of DEs from the Full Circumference Examination of the Inner Container Closure Weld Region (FY2020)

One of the main focus areas of the 3013 Surveillance Program is a thorough evaluation of the inner container closure weld region (ICCWR) opened for destructive examination (DE). As part of the protocol to investigate the corrosion in the ICCWR a laser confocal microscope (LCM) is used to perform close visual examination of the surface and measurements of corrosion features on the surface. DE containers from FY13 through FY16 were screened with three candidates selected for full circumference evaluation (FCE) according to the ICCWR examination protocol, During the FCE, FY16 DE05 and FY15 DE07 showed suspect major corrosion events, FY16 DE05 Section C2 shows two crack-like features, identified as Denebola and Draco. Both features are located at the boundary of Zone 2 and Zone 3. Consequently, Section C2 of FY16 DE05 was selected for examination by serial metallography at Savannah River National Laboratory (SRNL). For FY15 DE07 the suspect corrosion events were observed on Sections Cl and C2. Section Cl shows one crack-like feature, identified as Acrux. Section C2 shows three crack-like features, identified as Bellatrix, Cursa Minor and Cursa Major. Unlike Acrux and Bellatrix, which are located at the boundary of Zone 2 and Zone 3, Cursa Minor and Cursa Major are located in Zone 2. Sections Cl and C2 of FY15 DE07 were sent to Los Alamos National Laboratory (LANE) for characterization by X-Ray Tomography (XRT).

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

Composition Measurements of the HLW HAlG Glasses

This report provides the results from the chemical analyses of the glass compositions of the High-Level Waste High-Aluminum Glass study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with targeted compositions. The relative differences between the targeted and measured concentrations of Cr 2 O3, P 2 O 5 , and ZrO 2 for several of the glasses were greater than 10%. The relative difference between the targeted and measured concentrations of CaO was greater than 10% for one glass. The relative difference between the targeted and measured concentrations of Li 2 O was greater than 10% for one glass. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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

Grain boundary facilitated dissolution of nanocrystalline NpO 2 (s) from legacy waste processing

Dissolution of actinide dioxides, including neptunium dioxide (NpO 2 (s)), is paramount for the prediction of the environmental fate of nuclear materials. Quantifying dissolution rates, as well as understanding qualitative dissolution mechanisms, informs performance assessment for geologic disposal of spent nuclear fuel and management of legacy radioactive waste. The aim of this research was to measure the dissolution rate of nanocrystalline NpO 2 (s), produced through legacy nuclear waste processing, under oxidizing conditions, as well as to characterize surface alteration to the material. The solid phase was characterized using electron microscopy techniques (SEM/STEM) and X-ray photoelectron spectroscopy (XPS), indicating preferential dissolution of Np-hydroxide contained in the grain boundaries of NpO 2 (s) and fragmentation of grains from the matrix. The oxidative dissolution was monitored over 40 weeks, yielding a two-step kinetic dissolution model involving hydration of NpO 2 (s) and subsequent oxidation and dissolution of the hydroxide phase. Here, the proposed dissolution models for nanocrystalline NpO 2 (s) suggest that microstructural features such as grain boundaries are key factors affecting dissolution, including release of colloidal particles, and ultimately, environmental fate and transport of nuclear materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Characterization Strategy of a Limited Number of Radioactive Waste Effluent Tanks in View of Decommissioning a Large Number of Tanks - 20286

In order to organize the D and D of the former effluent treatment facility at the CEA Cadarache Center (France), the key data to gather through the facility characterization are the following:: volumes and masses, liquid/solid distribution and specific activity. By studying the treatment processes and the history of the facility, families of tanks have been defined and a reference tank chosen for investigation in each family. The intrusive investigations (video, residue thickness, dose rate..), the gamma spectrometry measurements carried out on site on the residue samples and radiochemical results collected in laboratories, led us to conclude that there was a relative radiological homogeneity (for specific activity and Am-241/Cs-137 ratio) within each tank. The next step consisted in extrapolating the specific activity evaluated for the reference tanks, to the other tanks of each family by using data such as geometry, dose rate or the function of the tanks within the process. Thus, a cartography of the 130 tanks of the facility is available. Finally, the conclusions drawn from the radiological homogeneity of the tanks and their affiliation to homogeneous, well-identified tank clusters lead to a simple and efficient operational approach for the D and D of the 130 tanks of the facility. Indeed, because of the residue characteristics (homogeneous activity and strong differentiation of Cs/Am ratio), a sample collected when opening the tank, and measured on site by gamma spectrometry will allow us to validate the operational design and the emptying techniques defined during the D and D study phase thanks to the characterization of a limited number of reference tanks. (authors)

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

A LEAN SYSTEMS APPROACH TO IMPROVING MAINTENANCE OUTPUT AND CONFIGURATION MANAGEMENT

The purpose of this capstone paper is to apply a lean systems approach to the current maintenance planning program in the Assembly/Disassembly Organization. This capstone will take a two-pronged approach to address both output rate in maintenance planning and accurate configuration management resulting from maintenance work orders. In short, the current system suffers from waste in the planning process due lack of procedures to address varying maintenance packages, including maintenance packages which include changes to configuration managed systems. The capstone will detail a lean systems approach to proper characterization of maintenance jobs and integration of change management into the maintenance process in order to retain proper control of the systems used to produce in this nuclear facility.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Characterization of Tank 9h Salt Dissolution Batch 2C In Support of Tank Closure Cesium Removal (TCCR) 1A Batch 2 Preparations

Savannah River Mission Completion (SRMC) is currently preparing the second batch of material to be processed through the Tank Closure Cesium Removal (TCCR) 1A system. The feed for TCCR 1A consists of dissolved saltcake from Tank 9H. The third batch of salt to make up processing Batch 2 (Batch 2C) has recently been dissolved in Tank 9H and transferred to Tank 10H where it was composited with the first part of the batch (Batches 2A and 2B) in preparation for processing through the TCCR 1A unit. Savannah River National Laboratory (SRNL) received samples from the recent batch (2C) of dissolved salt for characterization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Techno-economic feasibility of borehole thermal energy storage system connected to geothermal heat pumps for seasonal heating load of two buildings in Fairbanks, Alaska

Borehole thermal energy storage (BTES) is an effective solution for managing imbalanced heating and cooling loads in cold regions. This study evaluated the long-term feasibility of a BTES system in the Fairbanks area, Alaska, through building energy modeling, resource characterization, and numerical modeling. The system was designed to store waste heat from a nearby coal power plant during summer and provide thermal energy during winter to geothermal heat pumps supplying heating loads in two buildings. Heating load profiles were modeled for the buildings using EnergyPlus, and the results indicated the annual heating load was 5.6 times greater than the cooling load. 40 borehole heat exchangers were pre-designed approximately 100 m away from the two buildings in terms of land availability and regulatorily optimized depth. The 20-year performance of the designed BTES system under two operational scenarios—one with a 5-year preheating period and one without—was numerically modeled using subsurface temperature and properties characterized through the literature review and thermal response tests. Both scenarios demonstrated that the BTES has the capacity to fully cover the heating loads in the two buildings throughout the 20-year lifetime. Production temperatures at central wells were 33 % higher on average than at outer wells in both scenarios. The 5-year preheating period increased subsurface and extraction temperatures, and correspondingly annual average and total thermal energy production was higher for 8 years than in the scenario without the preheating period. These results highlight the long-term reliability and sustainability of the BTES system in meeting heating demands over its lifetime, with the preheating period offering potential performance improvements. Implementing the BTES system in cold regions with high heating demand, such as Fairbanks, Alaska, could provide a long-term, sustainable energy solution for managing imbalanced heating and cooling loads.

15 GEOTHERMAL ENERGY↗
DOE OSTI2024

Advanced Reactors Spent Fuel & Waste Science and Technology Program

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterization of Salt Waste Processing Facility August 2022 Tank 201 sample

Two Salt Waste Processing Facility (SWPF) Tank 201 samples were collected from the replacement of the decontaminated salt solution (DSS) coalescer at different times; one in December of 2021 and the other in August 2022. The December 2021 sample was collected after nitric and water flushing of the DSS Coalescer. The Aug 2022 sample was collected after water flushing only. The December 2021 Tank 201 solids were previously characterized and documented. SWPF requested the characterization of the August 2022 Tank 201 sample, and the comparison of the characterization results for these two Tank 201 solids collected at different dates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Summary of Results from November 2021 Qualification Samples for Tank Closure Cesium Removal 1A (TCCR 1A)

Savannah River Remediation (SRR) is currently operating the Tank Closure Cesium Removal 1A (TCCR 1A) process to remove 137 Cs from tank waste supernate using an ion exchange process. As part of that process, Savannah River National Laboratory (SRNL) receives and analyzes samples in support of the qualification of each batch to be processed. SRNL recently received supernate samples retrieved from Tank 10H as well as in-tank batch contact samples for characterization in support of qualifying Batch 1 for processing through the TCCR 1A unit.

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

Identification and Quantification of Technetium Species in Hanford Waste Tank AN-102

Technetium-99 (Tc) generated from the fission of 235U and 239Pu in high yields is one of the most difficult contaminants to be addressed at the U.S. Department of Energy Hanford Site. In strongly alkaline solutions typifying Hanford tank waste, Tc exists as pertechnetate (TcO4-) (oxidation state VII) as well as in reduced forms (oxidation state < VII) collectively known as non-pertechnetate species. Designing strategies for effective Tc management, including separation and immobilization, necessitates understanding the molecular structure of the non- pertechnetate species and their identification in the actual tank waste samples, which would facilitate development of new treatment technologies effective for dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO)3]+ and Tc(IV, VII) compounds was generated using a range of techniques and applied to the characterization of the actual tank waste supernatant collected from the tank 241-AN-102 at Hanford, WA. A sample of the 241-AN-102 tank waste supernatant was processed to adjust Na concentration to about 5.6 M and remove 137Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin. Cesium-loaded sRF column was eluted with 0.5 M HNO3. As received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using 99Tc nuclear magnetic resonance spectroscopy and X-ray absorption spectroscopy. It was demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO)3]+ and Tc(IV) compounds. This is the second experimental observation of the [fac-Tc(CO)3]+ species in the Hanford tank waste and the first demonstration of multiple forms of non-pertechnetate species existing simultaneously in the waste, cumulatively highlighting their importance for the waste processing.

Low Activity Waste (LAW), High Level Waste, nuclea↗
DOE OSTI2023

Overview of System Integration Analysis Activities for Integrated Waste Management

Spent nuclear fuel (SNF) generated by the current fleet of commercial nuclear reactors is being stored at the reactor sites in spent fuel pools (SFPs) and in dry independent spent fuel storage installations (ISFSIs). The US Department of Energy Office of Nuclear Energy (DOE-NE) is developing an Integrated Waste Management Program (IWMP) comprising a suite of options and supporting analyses to enable future informed choices. The IWMP is organized into the following five major areas: 1) Consent-Based Siting, 2) IWM Facilities and Equipment Concepts and Development, 3) Transportation Capability Analysis and Support, 4) Information Technology Solutions and Support, and 5) System Integration Analysis and Support. This paper discusses the activities ongoing in the IWMP System Integration Analysis and Support area. Two main areas of research in system integration are data and tools development, as well as system analysis assessments. One of the tools being developed in the system integration area is the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) tool. It is being developed as a foundational resource for DOE-NE to manage SNF data, along with several compatible analysis tools for time-dependent characterization of SNF and related systems. UNF-ST&DARDS has the unparalleled ability to track SNF through the entire back end of the fuel cycle—from the time the fuel is discharged from a reactor through its disposal in a geological repository. UNF ST&DARDS interfaces with the SCALE code system for nuclear analysis and COBRA-SFS for thermal analysis. Another main tool being developed is the Next Generation System Analysis Model (NGSAM). NGSAM is an agent-based simulation software tool expressly designed to be capable of modeling the waste management system. NGSAM has been developed to enable informed decision-making by providing the capability of analyzing various potential system options for the management of SNF and HLW. Using NGSAM, system architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). System analysis assessments may investigate the implications of various strategies such as different acceptance rates, acceptance queues, facility capacities and options, standardized canisters, and different assumed system operation start dates. Recently, some system analysis effort has begun to look at how the waste management system might operate for advanced reactor fuel cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

L-Basin Microbial Monitoring Program - Evaluation of 20 Years of Monitoring

The SRNL L-Basin corrosion surveillance and microbial monitoring programs provide early detection and characterization of corrosion attack to the fuel and storage system materials resulting from prolonged exposure to the L-Basin water environment and of changes to and impact of the diverse microbial population, respectively. The early detection of corrosion allows for adjustment of the water quality, engineering management, and fuel storage configurations to mitigate excessive corrosion attack. While microbial influenced corrosion in L-Basin has not been detected, tracking and understanding the effect of microbial populations on the stored fuel and basin water will aid in identifying remedial measures to mitigate any detrimental impact. This report reviews the microbial monitoring activities since initial characterizations in the mid-1990s.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗