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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

A History of Hanford Tank Waste, Implications for Waste Treatment and Disposal

More than 40 years of plutonium processing have left almost 56 million gallons of mixed radioactive waste sequestered in 177 underground tanks on the Hanford Site. Three different processing technologies were employed for plutonium purification in addition to uranium scavenging and fission product removal from the tank waste. All of these chemical processes have contributed to a complex waste stream that varies from tank to tank that presents downstream processing challenges to render the waste into a safe form for long-term storage. The current disposition pathway for Hanford tank waste is vitrification. To maximize waste loading and minimize the number of high-level waste canisters stored in a geologic repository, pretreatment of the waste is required. Both pretreatment and vitrification operations are impacted by the waste composition.

Separations, filtration, ion exchange↗
DOE OSTI2021

Saltstone Waste Characterization Analysis - Salt Waste Processing Facility (SWPF) Waste Stream (Q4 CY2020)

The Saltstone Production and Disposal Facility is designed and permitted by the state of South Carolina Department of Health and Environmental Control (SCDHEC) to treat and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LEW) aqueous streams from the Effluent Treatment Project (ETP) and decontaminated solutions from the Tank Closure Cesium Removal Unit (TCCR) and the Salt Waste Processing Facility (SWPF) are stored in Tank 50 until the LEW can be transferred to the Saltstone Facility for treatment and disposal. In the past, decontaminated solution from the Modular Caustic Side Solvent Extraction Unit (MCU) was stored in Tank 50 until the LEW could be transferred to the Saltstone Facility for treatment and disposal. MCU is currently in a suspended operations state. LEW that meets the Waste Acceptance Criteria (WAC) can be transferred, stored, and treated in the Saltstone Production Facility (SPF) for subsequent disposal as saltstone grout in the Saltstone Disposal Facility (SDF). Sampling will be conducted as new waste streams are identified for treatment and disposal at the Saltstone Industrial Wastewater Treatment Facility (IWTF) and Z-Area Industrial Solid Waste Landfill (ISWLF), Facility ID# 025500-1603, General Condition B.9 or every six years in accordance with South Carolina (SC) Regulation 61-107.19 Parti C, “Solid Waste Management: Solid Waste Landfills and Structural Fill - General Requirements.”

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

Waste Retrieval Enhancements to Achieve Preliminary Cease Waste Removal in Savannah River Site Liquid Waste Tanks 9H and 10H – 25348

The Liquid Waste (LW) contractor at the Savannah River Site (SRS) is Savannah River Mission Completion (SRMC). The LW Mission is tasked with processing legacy nuclear waste stored in underground waste tanks for final disposition. The Concentration, Storage, and Transfer Facilities (CSTF) contain 43 active waste tanks and 8 closed waste tanks between the two tank farms, F-Area Tank Farm (FTF) and H-Area Tank Farm (HTF). The first steps in the Waste Retrieval and Tank Closure (WRTC) process are the waste removal campaigns, consisting of either salt dissolution or sludge mobilization. Two tanks that are rapidly approaching the final closure determination and have demonstrated considerable success with salt dissolution are Tanks 9 and 10. The closure of these tanks is a high priority for the LW Mission due to the greater environmental risk they pose since both tanks reside within the water table and contain active leak sites from the primary tank to the annulus space. Tanks 9 and 10 have each recently completed their respective salt dissolution campaigns and achieved the Preliminary Cease Waste Removal (PCWR) milestone.

Stetson, Jacqueline G.↗
DOE OSTI2021

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the TRU Waste Processing Center Mixed Low Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) Transuranic (TRU) Waste Processing Center Mixed Low Level Waste (MLLW), FWORCHMLLW103, Revision 13 [TWPC 2021]), is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TRU Waste Processing Center MLLW meets all performance objectives of DOE Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The TRU Waste Processing Center MLLW waste stream is recommended for acceptance without conditions.

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

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Transuranic Waste Processing Center Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Transuranic Waste Processing Center (TWPC) Low-Level Waste (LLW), FWORCHLLW0102, Revision 12 (TWPC 2021), is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC LLW meets all performance objectives of DOE Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The TWPC LLW waste stream is recommended for acceptance without conditions.

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

Drying model of a high salt content cementitious waste form: Effect of capillary forces and salt solution

Highlights: • Drying model for a high salt content cementitious waste form is developed. • Water vapor diffusion and capillary liquid flow are distinguished. • Capillary and salt solution effects are considered in vapor-liquid equilibrium. A water transport model coupling capillary liquid flow with vapor diffusion is developed to describe the drying process for a cementitious waste form with high salinity porewater. Vapor-liquid equilibrium is formulated as the driving force for vapor diffusion and the model accounts for pore capillary and high salinity effects on water thermodynamic activity. Pore filling and porewater surface tension as a function of pore size distribution and water saturation have been quantified for the material. Geochemical speciation modeling is used to simulate porewater activity as a function of composition over the range of saturation. The theoretical relationship between relative humidity and water saturation generally agrees with experimental measurement, and the developed model is capable of predicting drying rates under various external relative humidity conditions. The model was developed to be incorporated into reactive transport models considering the effects of drying such as salt redistribution and efflorescence.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Life cycle analysis of polylactic acids from different wet waste feedstocks

Producing a valuable chemical product through diversion of wet wastes can simultaneously resolve the problems associated with increasing wastes and greenhouse gas emissions from conventional chemical production processes. In this work, we investigated the life-cycle greenhouse gas emissions, water, and fossil-fuel consumption for waste-derived polylactic acids (PLA) from three different waste feedstocks, namely wastewater sludge, food waste, and swine manure, using the Greenhouse Gases, Regulated Emissions, and Energy Use in Technologies (GREET) model. The decarbonization potential of replacing fossil-based resins with the waste-derived polymer was also investigated. The results show that swine manure-to-PLA pathway was the least carbon intensive (—1.4 kgCO 2 e/kg) among the three waste-to-PLA pathways on a cradle-to-grave basis, followed by the food waste case (—1.3 kgCO 2 e/kg) and then by the wastewater sludge case (0.6 kgCO 2 e/kg). In the baseline scenario, all three waste-to-PLA pathways were less carbon intensive than both fossil-based PET and HDPE on a cradle-to-grave basis: 66% (vs. PET) and 56% (vs. HDPE), 171 and 192%, 181 and 205% reduction in GHG emissions for wastewater sludge-, food waste-, and swine manure-to-PLA pathway, respectively. For all sensitivity cases investigated, the food waste- and swine manure-to-PLA pathways were significantly less carbon intensive than their fossil-counterparts. In terms of the annual decarbonization potential of replacing fossil-based PET or HDPE, the wastewater sludge- and food waste-pathway showed higher mitigation potential than the swine manure-pathway: i) 18–28 kilotons CO 2 e-reduction per year for wastewater sludge pathway; ii) 23–26 kTCO 2 e-reduction/yr for food waste pathway; and iii) about 5 kTCO 2 e-reduction/yr for swine manure pathway depending on the type of conventional resin replaced. However, given the abundant availability of the swine manure feedstocks across the United States, the decarbonization potential of swine manure-based pathway can also increase as the plant capacity or the number of plants grow.

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

Modeling delayed thermal runaway in nitric acid-soaked cat litter mixed with radioactive waste

Thermal ignition of radioactive waste within a 55-gallon drum was simulated by using a pressure-dependent waste decomposition model (Hobbs et al. in Process Saf Environ Prot https://doi.org/10.1016/j.psep.2022.09.047, 2022) calibrated with data from full-scale drum experiments (Parker et al. in The thermolytic response of a surrogate RNS waste mixture at the drum scale. Los Alamos National Laboratory Report LA-UR-16-21760, 2016) and validated with experiments from multiple laboratories (Hobbs et al. in Thermal analysis of aged nitric acid-soaked kitty litter in TRU waste drums-23370.WM2023 Conference, Phoenix, AZ, 2023). The acceleration of nitric acid chemistry reacting with an organic cat litter leading to thermal ignition was likely triggered by a restricted vent in the drum. Here, we address whether the form of the rate equation in (Hobbs et al. in Process Saf Environ Prot https://doi.org/10.1016/j.psep.2022.09.047, 2022) is sufficient to extrapolate thermal ignition within aged drums of similar content that have been stored in Texas for over nine years by investigating four different reaction rate forms for waste decomposition. A critical reaction rate reduction analysis is performed on each of these models to determine if delayed thermal runaway within vented aged waste is possible after nine years. We found that a pressure-dependent first-order rate expression not only predicted the accidental ignition of the waste drum, but the form also matches multiple experiments from different laboratories. Even though the waste composition decreases over time, the model predicts that acceleration leading to thermal runaway is possible if the waste is confined, even after 9 years. In conclusion, waste containing oxidizers such as nitric acid should not be mixed with organic adsorbents, especially if the waste is confined.

cookoff↗
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 OSTI2020

Self-accelerated corrosion of nuclear waste forms at material interfaces

The US plan for high-level nuclear waste includes the immobilization of long-lived radionuclides in glass or ceramic waste forms in stainless-steel canisters for disposal in deep geological repositories. Here we report that, under simulated repository conditions, corrosion could be significantly accelerated at the interfaces of different barrier materials, which has not been considered in the current safety and performance assessment models. Severe localized corrosion was found at the interfaces between stainless steel and a model nuclear waste glass and between stainless steel and a ceramic waste form. The accelerated corrosion can be attributed to changes of solution chemistry and local acidity/alkalinity within a confined space, which significantly alter the corrosion of both the waste-form materials and the metallic canisters. The corrosion that is accelerated by the interface interaction between dissimilar materials could profoundly impact the service life of the nuclear waste packages, which, therefore, should be carefully considered when evaluating the performance of waste forms and their packages. Moreover, compatible barriers should be selected to further optimize the performance of the geological repository system.

Guo, Xiaolei↗
DOE OSTI2020

Waste Information Management System with 2019-20 Waste Streams - 20491

The Waste Information Management System (WIMS) 2019-20 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 35 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and intermodal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)

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

Advanced Mixed Waste Treatment Project Overcomes Challenges to Complete Treatment of Debris Waste - 20616

The Advanced Mixed Waste Treatment Project (AMWTP), operated by Fluor Idaho for the Department of Energy Idaho Operations Office, completed the last of its debris waste treatment in the fall of 2019. Concluding that work-scope was certainly not without its challenges. Fluor Idaho and its employees had to adapt to treating extremely challenging waste types, such as high-fissile gram materials and potentially pyrophoric and reactive wastes coming down the stretch. It also had to complete the sizing of large waste boxes from Mound Laboratories in inner-contamination enclosures designed to capture the extremely fine, 'flighty' plutonium-238 in its high-efficiency particulate air filters. Very few waste treatment facilities in the DOE Complex have completed their missions. In fact, most large-scale waste treatment facilities are still under construction with many months or years remaining before 'hot operations' begin. AMWTP is unique because it functioned as a 'one-stop shop.' Transuranic and low-level waste characterization, treatment, certification, and shipping all occurred there. AMWTP concluded the most challenging part of its mission and will continue to ship material out of the state of Idaho for the next decade. This paper will provide a background of the AMWTP mission, and the adaptations to its operating permit, equipment, or treatment processes to address challenging waste types. AMWTP will leave behind a legacy of being one of the most successful waste treatment facilities in the history of the DOE Complex. The facility was well-conceived, well-constructed, and well-executed; it was done right from start to finish. During the course of its waste treatment mission, the facility was required to adapt to changing conditions by treating extremely challenging waste types. (authors)

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

Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form

This study demonstrates the conversion of simulated pyroprocessing salt wastes [i.e., ER(SF) and ERV2] into chemically durable iron phosphate glass waste forms through reactions with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)2HPO4] and added Fe2O3. The reaction of the salt simulants with the phosphate precursors resulted in dehalogenation of the salt through the production of NH4Cl and NH4I (along with water), of which the NH4Cl can be used to produce UCl3 oxidant for use in the electrorefiner during pyroprocessing. These iron phosphate glasses contain a notably higher waste loading than the baseline glass-bonded sodalite ceramic waste form, which requires about 70 vol% sodalite to accommodate Cl from the salt, because they are dehalogenated prior to vitrification. The high waste loadings coupled with higher bulk densities (~3×103 kg/m3) than glass-bonded sodalite (~2.3×103 kg/m3) mean that these glasses have a higher storage volume, i.e., the volume required to immobilize a given mass of waste. The iron phosphate glass waste forms were characterized using a variety of techniques including physical property measurements and chemical durability testing. Following initial testing, a proof-of-concept demonstration was conducted to show that the NH4Cl generated from this process could be used to produce UCl3 using depleted uranium dendrites as a feedstock.

phosphate glass, pyroprocessing, waste forms, salt↗