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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
NASA NTRS1980

Space disposal of nuclear wastes

The DOE has been studying several options for nuclear waste disposal, among them space disposal, which NASA has been assessing. Attention is given to space disposal destinations noting that a circular heliocentric orbit about halfway between Earth and Venus is the reference option in space disposal studies. Discussion also covers the waste form, showing that parameters to be considered include high waste loading, high thermal conductivity, thermochemical stability, resistance to leaching, fabrication, resistance to oxidation and to thermal shock. Finally, the Space Shuttle nuclear waste disposal mission profile is presented.

Priest, C. C.↗
DOE OSTI2016

Inorganic tin aluminophosphate nanocomposite for reductive separation of pertechnetate

Pertechnetate (TcO4-) is the most abundant chemical form of radioactive contaminant 99Tc present in the legacy nuclear waste streams and in the subsurface of the nuclear waste storage sites.One proposed remediation approach is reductive separation of TcO4-and sequestration in low-temperature waste forms. The development of the relevant technologies has been slow due to the lack of reductive materials that retain their functionalityand are otherwise suitable for application in multicomponent or aggressive media such as highly alkaline, brine-like solutions typifying nuclear tank wastes. This research prepared a tin-based reductive material and demonstrated its potential utility for the separation of TcO4- from the alkaline nuclear wastes. This material consists of Sn(II/IV) phosphate supported by the polycrystalline aluminophosphate matrix. The aluminophosphate matrix is inert to the reaction conditions and offers the benefits of high stability and low solubility in the concentrated alkaline solutions. This Sn(II/IV)-basedmaterial exhibitshigh loading capacity for Tc and selectively removed major fraction of TcO4-from the tank waste supernatant simulants, which contains 7.8 M total sodium and 2.43 M free hydroxide concentrations. The observed Kd values for Tc are about 13,000 and 2,200 mL g-1 for the simulant solutions containing no or 33 mM Cr(VI), respectively, positioning Sn(II/IV) aluminophosphate among the best performing reductive sorbents for TcO4- developed to date. This advanced behaviour is attributed to the synergistic combination of the Sn(II/IV) aluminophosphate functionalities. The presence of Sn(II/IV)-rich fibres facilitates the reduction of TcO4- to Tc(IV), which getsimbedded along the fibre branches. Importantly, the Sn(IV)-containing inert polycrystalline matrix alsoincorporates Tc(IV)which triggers its crystallization to rutile SnO2 phase andstabilizesTc(IV) in the polycrystalline matrix.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2020

Composite Analysis Solid Waste Release Data Reduction of Activity Flux from Waste Sites to the Vadose Zone for Baseline Assessment

This environmental calculation file (ECF) describes the data reduction methodology for the solid waste release datasets generated by the Composite Analysis Solid Waste Release (CASWR) model (CP-62766, Model Package Report: Composite Analysis Solid Waste Release Model [CASWR Model]) and documents the resulting reduced datasets for the updated Composite Analysis (CA) baseline assessment. This ECF is in support of the updated Hanford Site CA. The purpose of the updated Hanford Site CA is to provide an estimate of the cumulative radiological impacts from active and planned low-level radioactive waste disposals and other potentially interacting radioactive waste disposal sources that will remain following Hanford Site closure. The vadose zone modeling activities associated with the CA includes solid waste forms as contaminant sources. To support the modeling of solid waste transport from the vadose zone to the groundwater at the Hanford Site, the CASWR model was designed to generate deterministic radionuclide release rates for the Hanford Site Central Plateau solid waste disposal sites. The complete CASWR model-generated dataset for the CA Baseline Assessment consists of 2,378,831 sets of time and release rates (1,073,869 sets for the 200 East Area and 1,304,962 sets for the 200 West Area), representing a course of the simulation from year 1953 to year 12,069, 16 radiological contaminants of potential concern (COPCs), and 237 solid waste sources as documented in ECFHANFORD- 19-0112, Solid Waste Release Calculations for the Composite Analysis Baseline Assessment. For several of the STOMP vadose zone models, the number of release rate data pairs required for release at all waste sites would exceed the limitations imposed by the STOMP modelling software package used to simulate contaminant transport through the vadose zone. As a result, a data reduction of the CASWR dataset is required prior to its use as a source input for the CA vadose zone modeling activities.

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

Affinity of LDR Organics to Cementitious Materials and Activated Carbon

The Hanford site stores approximately 56 million gallons of radioactive and hazardous waste in underground storage tanks. The current low-activity waste (LAW) vitrification facility does not have the capacity to immobilize all of the LAW. Therefore, cementitious waste forms are being evaluated as a cost-effective supplemental waste treatment, focusing on the solidification/immobilization of Land Disposal Restricted (LDR) organics, which is currently not recognized by the Environmental Protection Agency (EPA) as a standard treatment method (40 CFR 268.42). Previous efforts by the EPA and waste management agencies have highlighted the potential for cementitious materials to retain certain organic species through physical (encapsulation) and chemical (sorption) interactions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Stabilization of Radionuclides and Metals into Spinels Using THOR{sup R} Fluidized Bed Steam Reforming - 12039

Studsvik, Inc.'s Thermal Organic Reduction (THOR) Process has been operating for over 12 years in the treatment of radioactive liquid and solid wastes from commercial Nuclear Power Plants in the United States. A second THOR plant is starting operations in early 2012 for the treatment of radioactive wastes from US nuclear power research at US DOE's Idaho National Laboratory. THOR has long been recognized for its unique ability to substantially reduce the volume of radioactive wastes - via the removal of water and destruction of organics, nitrates/nitrites, etc.- and produce a dry, stable solid product while retaining extremely high amounts (>99.99%) of the incoming wastes' radioactivity. Moreover, the THOR process has demonstrated the ability to produce radioactive waste forms with leach-resistance as good as or better than vitrified glass. An additional facet of the technology which, to this point, has been marginalized is the production spinels that are essential for capturing volatile, leachable, and large radioactive and hazardous inorganic species. These spinels form from the wide variety of inorganics in the incoming waste feed and proprietary additives used in processing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Structural Origins of High MoO 3 Solubility in Peraluminous Borosilicate Glasses

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

Amorphous materials↗
DOE OSTI2017

Glass-ceramics for nuclear-waste immobilization

Crystallization in glasses is usually considered to be a problem in the glass industry. However, controlled crystallization of glasses is an important prerequisite in the development of glass-ceramics with tailored useful properties. Similar boundary conditions apply when considering glass-ceramics for the immobilization of nuclear waste via vitrification. While uncontrolled crystallization in nuclear-waste glasses is problematic, chemically durable glass-ceramics with significantly high waste loadings can be produced with controlled crystallization of glasses. This article presents an overview of various aspects of nuclear-waste glasses where crystallization is either considered to be advantageous or problematic. The classification of glass-ceramic waste forms and strategies to design glass-ceramics for a given waste stream is discussed. In conclusion, some open and relevant problems faced by researchers developing nuclear-waste glass-ceramics are also offered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

Cold crucible induction melter↗
DOE OSTI2018

2017 NE-4 Nuclear Technologies Research and Development (NTRD) Achievements Report

Fiscal year (FY) 2017 initiated new and reestablished presidential and departmental nuclear energy priorities for the U.S. Department of Energy (DOE), Office of Nuclear Energy (DOE-NE) that will: • Support the Administration’s priorities to achieve U.S. energy dominance and economic prosperity through U.S. technology leadership • Support DOE-NE mission priorities and goals for the existing nuclear fleet, advanced reactor pipeline and national fuel cycle infrastructure • Focus on early stage research, and effective cooperation with industry, academia, and other government agencies • Propose investments to maintain and improve capabilities and infrastructure to conduct leading-edge science, research, and technology development. DOE-NE’s mission priorities will focus on the existing nuclear fleet, advanced reactor pipeline, and fuel cycle infrastructure. Therefore, the DOE Nuclear Technology Research and Development (NE-4) mission aligned to (1) support the research and development (R&D) of innovative reactor technologies that may offer improved safety, functionality, and affordability and (2) conduct research to reduce long-term technical barriers for advanced nuclear energy systems, including space fission systems. Key activities to execute this mission include: • To achieve the NE-4 mission priorities, two major programs with eight R&D campaigns are engaged in impactful research, development, demonstration, and deployment: Fuel Cycle Technologies (FCT) Program • Fuel Cycle Options Campaign evaluates complex fuel cycle options and provides performance metrics on various systems and scenarios to improve the knowledge of interdependencies between technologies and different systems. • Advanced Fuels Campaign performs science-based R&D on accident tolerant fuels in addition to advanced light-water reactor (LWR) fuel and transmutation fuel. • Joint Fuel Cycle Studies (JFCS) is a collaboration with the Republic of Korea to assess the feasibility and nonproliferation of electrochemical recycling to manage used fuel. • Material Recovery and Waste Form Development (MRWFD) Campaign researches advanced fuel cycle material recovery and waste management capabilities to improve fuel cycle performance with less processing and waste generation. • Material Protection, Control, and Accountability Technologies (MPACT) Program develops tools and capabilities to secure the next generation nuclear materials management and safeguards for nuclear fuel cycles. Advanced Reactor Technologies (ART) Program • Fast Reactors Campaign targets benefits in either resource utilization or waste management and possesses favorable features for small reactor applications and plutonium management. • Gas-Cooled Reactors Campaign seeks to raise the technological readiness of high temperature gas-cooled reactors (HTGRs) and other reactor concepts utilizing coated particle fuel, graphite, or helium coolant by performing high risk, enabling research and development that commercial developers are unable or unwilling to undertake in the current energy market. • ART Energy Conversion (EC) Project provides solutions to convert heat from an advanced reactor to useful products that support commercial application of the reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

PNNL Supports Hanford Waste Treatment

For more than 40 years, technical assistance from PNNL has supported the operations and processing of Hanford tank waste. Our expertise in tank waste chemistry, fluid dynamics and scaling, waste forms, and safety bases has helped to shape the site’s waste treatment baseline and solve operational challenges. The historical knowledge and unique scientific and technical expertise at PNNL are essential to the success of the Hanford mission.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Aging Infrastructure: Application of System Health Monitoring at the Defense Waste Processing Facility - 16069

The disposition of high level waste (HLW) into borosilicate glass canistered waste forms began at SRS in 1996. Since that time, the Defense Waste Processing Facility (DWPF) has produced nearly four thousand HLW canisters. The DWPF is now approaching 20 years of radioactive service and 25 years of operation since the commencement of nonradioactive testing in 1990. Many facility systems are challenged by equipment or component obsolescence, which places a premium on proper work planning, engineering, and maintenance resource allocation to ensure facility reliability. To this end, Savannah River Remediation (SRR) implements programs and procedures for performance monitoring of structures, systems, and components (SSCs) within all of its processing facilities. As these programs have matured, they have helped to identify vulnerabilities within facility systems and provided a framework for ranking system assets to ensure correct priority is established and maintained. This paper provides an overview of how SRR structures and implements system performance monitoring with specific focus on the recent replacement of the DWPF Safety Grade Purge Flowmeters. This case study will discuss: 1) How the performance monitoring process identified an availability risk due to the minimal vendor support for the in-use flow transmitters, 2) The design changes developed as part of the project to transition to a different measurement technology, which includes operational lessons learned from the original installation, 3) The execution of the replacement effort, which was complicated by the fact that the purge system is a Vital Safety System (VSS) and not easily taken out of service for extended maintenance, and 4) The need to consider future maintainability, particularly given that continued acceleration in technical innovation will likely increase the frequency of equipment obsolescence. (authors)

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

Reactions during conversion of simplified low-activity waste glass feeds

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

Jin, Tongan↗
DOE OSTI2022

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

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

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

Solidification of REDC Organics For Disposal As Solid Waste

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

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Degradation of Simulated HLW Glass and International Simple Glass immersed partially in water - 16474

Water degradation of glass waste forms destined for geological disposal has been studied in a variety of conditions including of bulk glass immersed completely in static or dynamic water. During the course of repository life, the canisters may be only partially immersed in the initial stages of contact with water. Under these conditions glass degradation occurs on the glass immersed in water and on the glass above water by vapor condensation. In this work degradation of International Simple Glass (ISG) and Simulated High Level Waste Glass (HSG) is examined by immersing partially in distilled water during the course of test. Under static experimental conditions at 90 deg. C, the condensation of vapor on the ISG glass surface formed Ca-rich precipitates and pits after degrading for 1 and 3 weeks respectively. Whereas HSG precipitated Ca-rich phases predominantly after 1 week and no pits were found on the surface. When compared, HSG was found to change surface chemistry faster than ISG. (authors)

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

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

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

36 MATERIALS SCIENCE↗