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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 559 records · Page 31
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Solidification/Stabilization of Liquid Oil Waste in Meta-kaolin-based geo-polymer - 17209

Cement-base solidification/stabilization (S/S) is a widely applied technique for the immobilization of inorganic hazardous constituents. However, S/S is considered less compatible with organic liquid wastes (OLW) because organic compounds may inhibit binder hydration. geo-polymers (alumino-silicate materials) are a worthwhile material for applications in which cements are inappropriate. In this study, we propose a direct immobilization process of organic liquid waste in a geo-polymer matrix based on the emulsification of OLW in an activating solution, followed by the addition of aluminosilicate source (meta-kaolin). At first, the harmlessness of the chemicals contained in the OLW on the geo-polymerization was examined through a screening, reveling that only acidic oils should be pre-treated before a direct immobilization. Waste form characterizations with a simulated OLW were measured, showing interesting mechanical and confinement properties. Waste forms containing 20% Vol. of OLW show a mechanical strength higher than 30 MPa in compression and less than 1 % of waste leached over a month in pure water. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Novel inorganic sorbent for iodine removal in pump and treat system - 15020

The radioisotope {sup 129}I is one of the key risk components in the migrating subsurface plumes present in the Hanford Site vadose zone due to its long half-life, large inventory, high toxicity, and high mobility in the aquifer. Removal of iodine from groundwater in the 200 Area is an essential component for managing the Hanford Site cleanup and protecting the Columbia River. In the subsurface aquatic environment it mainly exist as iodide, iodate, and organic derivatives of iodine. Because of different chemical properties of these species and significantly lower concentration than other constituents in the subsurface plumes their removal using commercial ion exchange resins is problematic. In addition, availability of a stable matrix for a long-term waste form for iodine is needed. Novel engineered inorganic composites that possess superior characteristics including large surface area, high sorption capacity, and good chemical-mechanical-thermal stability may potentially offer a solution. These two-in-one novel materials can be designed to serve as an efficient sorbent for the simultaneous removal of the dissimilar iodine species and a durable waste form under environmentally relevant conditions. This paper describes our current results on identification, characterization, and testing of the iodine-selective inorganic sorbents. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Radiation Stability of Spark‐Plasma‐Sintered Lead Vanadate Iodoapatite

Spark‐plasma‐sintered lead vanadate iodoapatite Pb 9.85 (VO 4 ) 6 I 1.7 , a promising nuclear waste form for the immobilization of I‐129, was irradiated with energetic ions, electrons, and gamma rays, to investigate its radiation stability. In situ TEM observation of the 1 MeV Kr 2+ irradiation shows that lead vanadate iodoapatite generally exhibits higher tolerance against ion irradiation‐induced amorphization than lead vanadate fluorapatite, and the spark plasma sintering can further enhance its radiation stability attributed to the enhanced crystallinity, reduced defect concentration, and denser microstructure. The critical amorphization dose and critical temperature for the SPS‐densified iodoapatite at 700°C are determined to be 0.25 dpa at room temperature and 230°C, respectively. No significant phase transformation or microstructural damage occurred under energetic electron and gamma irradiations. Raman spectra of gamma‐ray‐irradiated iodoapatite indicate improved V–O bond order at 500 kGy dose. Generally, the spark‐plasma‐sintered iodoapatite exhibits excellent radiation stability for nuclear waste form applications. The significantly enhanced radiation stability of the SPS‐densified iodoapatite suggests that SPS holds great promise for fabricating iodoapatite waste form with minimum iodine loss and optimized radiation tolerance for effective management of highly volatile I‐129.

Lu, Fengyuan↗
DOE OSTI2018

Atomistic scale investigation of cation ordering and phase stability in Cs-substituted Ba 1.33 Zn 1.33 Ti 6.67 O 16 , Ba 1.33 Ga 2.66 Ti 5.67 O 16 and Ba 1.33 Al 2.66 Ti 5.33 O 16 hollandite

The titanate-based hollandite structure is proposed as an effective ceramic waste form for Cs-immobilization. In this study, quantum-mechanical calculations were used to quantify the impact of A-site and B-site ordering on the structural stability of hollandite with compositions Ba x Cs y (M z Ti 8-z )O16, where M = Zn 2+ , Ga 3+ , and Al 3+ . The calculated enthalpy of formation agrees with experimental measurements of related hollandite phases from melt solution calorimetry. Ground state geometry optimizations show that, for intermediate compositions (e.g., CsBaGa 6 Ti 18 O 48 ), the presence of both Cs and Ba in the A-site tunnels is not energetically favored. However, the decay heat generated during storage of the Cs-containing waste form may overcome the energetics of Ba and Cs mixing in the tunnel structure of hollandite. The ability of the hollandite structure to accommodate the radioparagenesis of Cs to Ba is critical for long term performance of the waste. For the first time, B-site ordering was observed along the tunnel direction ([001] zone axis) for the Ga-hollandite compositions, as well as the intermediate Al-hollandite composition. These compositionally dependent structural features, and associated formation enthalpies, are of importance to the stability and radiation damage tolerance of ceramic waste forms.

36 MATERIALS SCIENCE↗
DOE OSTI2013

The need for a characteristics-based approach to radioactive waste classification as informed by advanced nuclear fuel cycles using the fuel-cycle integration and tradeoffs (FIT) model

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. Because heat generation is generally the most important factor limiting geological repository areal loading, this analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. Waste streams generated in different fuel cycles and their possible classification based on the current U.S. framework and international standards are discussed. It is shown that the effects of separating waste streams are neglected under a source-based radioactive waste classification system. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Dismantling of the Spent Resins Hot Compaction Unit at Tihange NPP, Belgium - 19139

Design, construction and commissioning of the Spent Ion Exchange Resins Hot Compaction Unit at Tihange Nuclear Power Plant took place from 2005 till 2011. In this conditioning process, batches of radioactive resin beads were first dried to less than 4% moisture by a conical vacuum dryer. Resins were then poured in compactable 200-liter drums and super-compacted. Resulting pellets were emplaced in standard 400-liter drums, the packages qualified for final disposal in Belgium, and eventually grouted. The Belgian Waste Management Agency (ONDRAF/NIRAS) was consulted before initiating the project and issued a statement of no objection. The technology turned out to be used for the first time in the particular case of bead shaped resins (unlike powdered resins) and with a grouting requirement. Finally, commissioning tests ended in a safe and reproducible waste form, compliant with prevailing Waste Acceptance Criteria. Volume Reduction Factors of about 2 were recorded after drying and compaction. New tests were carried out to support the Qualification File of the process, which was submitted to ONDRAF/NIRAS. After performing many trials using inactive resins, the Operator dried two batches of radioactive resins and super-compacted four drums: this was a planned qualification step, aiming at the validation of process parameters. In the meantime, ONDRAF/NIRAS made parallel progress in the design of the future Near Surface Repository, where conditioned resins will be disposed of, and found some incompatibility between encapsulated dried resin beads and water intrusion hazards inside the repository. Based on the risk that dried resin beads will swell if they absorb water, causing mechanical stresses likely to compromise the integrity of the disposal site, ONDRAF/NIRAS informed Tihange NPP that the Hot Compaction Unit would not be qualified as spent resins conditioning process. Further to this irrevocable decision, the Operator undertook three actions: - The first one was the qualification and purchase of waste containers to provide additional temporary storage capacity on site for its operational spent resins; - The second one is the selection of a new spent resins conditioning process and the performance of the related qualification tests to demonstrate it meets the updated Waste Acceptance Criteria; - The last one is the dismantling of the Hot Compaction Unit to clear the space and enable the installation of new equipment. This latter task was carried out in 2018, involving decontamination and free release measurements of all the materials, a very limited fraction of it having to be further processed as radioactive waste. Nevertheless, Tihange NPP wishes to keep the super-compactor and its hydraulic unit for a possible resale or reuse for the treatment of mixed waste. (authors)

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

Special Analyses for the Hanford Integrated Disposal Facility Performance Assessment - 20102

In 2014, the Department of Energy (DOE) Office of River Protection and its contractors began to develop a performance assessment for the near-surface disposal of low-level and mixed low-level waste at the Hanford Site's Integrated Disposal Facility (IDF). The IDF is a doubly-lined landfill that was constructed between 2004 and 2006 to be the disposal facility for the vitrified low-activity waste that will be produced at the Waste Treatment and Immobilization Plant (WTP). IDF is also expected to receive solid secondary waste produced at the WTP and other solid wastes from site activities. The IDF has been in a preoperational state awaiting authorization from DOE and a RCRA permit modification from the State of Washington Department of Ecology to receive waste. Both the Disposal Authorization Statement and permit modification require a performance assessment demonstrating that the system of engineered and natural features will limit releases of radionuclides and hazardous chemicals from the IDF and be protective of human health and the environment. The simulated duration is 10,000 years. Based on the analyses presented in the 2017 Integrated Disposal Facility Performance Assessment, DOE issued a conditional Operating Disposal Authorization Statement for the IDF in June 2018. The long-term performance of the IDF to be protective of human health and the environment was evaluated under the requirements of DOE Order 435.1, Radioactive Waste Management. Computer simulations were performed to evaluate whether or not the IDF would comply with DOE requirements. In the time that has passed since the performance assessment was approved by DOE, new information has been discovered that had not been considered in the performance assessment. Since this new information has not been evaluated, the potential impact of the changes have not been taken into consideration in DoE's disposal authorization. DOE and its contractors follow a change control process to screen and, when necessary, evaluate new information that could potentially impact the conclusions of the completed performance assessment. This paper will describe the change control process and provide two examples of evaluations performed following the change control process. The first example evaluates a new waste form for liquid secondary waste that was not evaluated in the performance assessment. In the performance assessment, liquid secondary waste was assumed to be solidified with grout. A new recommendation to dispose of the liquid secondary waste after drying it to a powder was evaluated. The second example evaluates inventory implications from changes to the flow sheet that estimates the feed composition to the low-activity waste vitrification facility. The changes result in higher strontium concentrations in the vitrified waste stream. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Technetium incorporation in glass for the Hanford Tank Waste treatment and immobilization plant - 15066

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of {sup 99}Tc (∼9 x 10 E2 TBq or ∼2.5 x 10{sup 4} Ci or ∼1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with {sup 99}Tc is its high mobility in addition to a long half-life (2.1 x 10{sup 5} yrs). The pertechnetate ion (TcO{sup 4-}) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Infiltration Data Package for the E-Area Low-Level Waste Facility Performance Assessment

The infiltration data package contains the input parameters, the cap design and material properties assumptions, and the modeling results for the Hydrologic Evaluation of Landfill Performance (HELP) infiltration model simulations performed in support of the E-Area Low-Level Waste Facility Performance Assessment (PA). The infiltration estimates establish the upper boundary condition for the PORFLOW vadose-zone model and GoldSim model simulations for the following E-Area disposal unit types: Slit and Engineered Trenches, Low-Activity Waste Vault (LAWV), Intermediate-Level Vault (ILV), Componentin- Grout (CIG) special waste form trench segments, and the Naval Reactor Component Disposal Areas (NRCDAs). The infiltration data package builds upon relevant, foundational PA technical reports and memoranda from the past 15 years and is supported by three important components: the HELP model input parameter datasheets, HELP model input and output filenames and directory structure, and infiltration rates as a function of time for each scenario for each disposal unit type. For Slit and Engineered Trenches, a single best-estimate Bahia-grass case of two-percent slope and 585- foot slope length represents an upper bound on intact closure-cap infiltration rates over a 10,000-year period. Building upon the single, bounding, and intact case, two different approaches were considered for incorporating the effect of localized cap subsidence on infiltration rates in the PORFLOW flow and transport model simulations. In the first approach, a Monte Carlo probabilistic model was developed to calculate infiltration rates for 0.54%, 2%, 3.6%, and 4.9% subsidence scenariosa reflecting historical and future non-crushable content for the Slit and Engineered Trenches. The resulting blended (spatially and slope-length averaged) infiltration rates become an inflow upper boundary condition for the PORFLOW vadose zone model simulations. In the second and preferred approach, a Monte Carlo rectangle packing algorithm (Danielson, 2019) and a simplified equation for the total mass flux of water draining into a subsided compartment (Equation 1) will be employed to implement a weighted blending of radionuclide fluxes to the water table using the results of deterministic PORFLOW vadose zone simulations representative of specific subsidence cases. Finally, HELP-model results from sensitivity studies of the intact infiltration case were fit to a log-logistic function to generate infiltration profiles over a 10,000-year period for most-optimistic, more-optimistic, best-estimate, more-pessimistic, and most-pessimistic cases, for both the intact and four subsidence scenarios. For the LAWV and IL vaults, infiltration rate profiles for a 10,000-year period were developed for both an on-vault (above the concrete/metal vault roof) and an off-vault (10-foot soil zone adjacent to vault walls) scenario. The purpose of the off-vault simulations was to confirm that subsurface runoff from the concrete vault roof will adequately drain through the lowermost backfill layers adjacent to the vault walls. The LAWV and ILV infiltration profiles assume collapse of the concrete roof at relative Years 2,905 and 7,100, respectively. The identical Bahia-grass case with two-percent slope and 585-foot slope length assumed for the Slit and Engineered Trenches will represent an upper bound on intact infiltration rates for the NRCDAs and CIG special waste form as well. While the NRCDA casks are assumed to remain structurally intact (i.e., non-assumed to collapse 300 years after the end of operations. Because both rows of the existing CIG trench segments in ST23 (formerly CIG01) extend across the planned cap crest, the subsided CIG trench segments are treated conservatively in the HELP model as a 100%-subsidence case with zero subsurface run-on. The CIG special waste form trench segment infiltration profile, therefore, consists of an intact-cap period from relative Year 100 to Year 300, followed by a 100%-subsidence period from relative Year 300 to Year 10,100.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

DOE nuclear separation technologies - 15148

Realizing that challenges are facing US actinide science, separations, and waste form research and development activities including meeting future manpower needs, the U.S Department of Energy (DOE), Office of Environmental Management (EM), Office of Nuclear Energy (NE), and National Nuclear Security Administration (NNSA) jointly co-sponsored a workshop entitled 'Separations Technologies: 'Going from Where We Are to Where We Want to Be' on nuclear separations technologies in Bethesda, Maryland (July 2011) to identify common needs and potential requirements in separations technologies and opportunities for program partnerships. Recognizing the different mission requirements of the sponsoring programs, participants were able to identify areas with common uses and related applications. The two follow-on technical workshops held during the Fall 2012 addressed both the chemical aspects and the physical aspects of nuclear separations. Each workshop consisted of technical discussions regarding future work relevant to DOE goals. The large number of subject matter experts from the national laboratories, academia, and headquarters staff developed a long list of existing and new technology developments in the areas of aqueous separations, pyrochemical processing, waste forms, fuel technology etc... Each of these workshops identified the current state of knowledge or technology, the resources needed, technology development opportunities, and the objectives to reach the desired end-state(s). A consolidated report with its eight appendices will be issued at the end of December 2014 summarizing the outcomes of these workshops. Some examples of these findings will be presented in our contribution. It is clear that, in the current environment of decreasing budgets it is illogical for NE, EM and NNSA to develop its own home-grown technology. Therefore, there is a strong case to be made for leveraging limited resources in order to accomplish all objectives while significantly reducing development costs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Monte Carlo Simulation of Borosilicate Glass Dissolution using Molecular Dynamics-Generated Glass Structures

Atomic-scale insights into the dissolution mechanisms of borosilicate glasses are critical to the development of the waste form corrosion models that inform performance assessments of nuclear waste disposal facilities. Monte Carlo (MC) simulations have been instrumental in advancing the state of knowledge of borosilicate glass dissolution; however, all such simulations to date have relied on lattices to represent the amorphous glass structure, thereby potentially introducing artifacts. Therefore, a new approach, referred to as the amorphous MC approach, has been developed whereby glass structures generated from molecular dynamics simulations are used as starting points for MC simulations of glass dissolution. MC simulations of sodium borosilicate glasses that covered a wide compositional range were performed and revealed correlations between dissolution rate and structural features of the glasses. Notably, the lattice MC approach predicted the fraction of tetrahedral boron to have a lesser influence on the dissolution rate than computed with the amorphous MC approach. The lattice MC simulations also overestimated the dissolution rate for a given glass composition. These findings highlight the importance of using truly amorphous structures in MC simulations of glass corrosion.

Kerisit, Sebastien N.↗
DOE OSTI2018

Compositional Dependence of Solubility/Retention of Molybdenum Oxides in Aluminoborosilicate-Based Model Nuclear Waste Glasses

We present that molybdenum oxides are an integral component of the high-level waste streams being generated from the nuclear reactors in several countries. Although borosilicate glass has been chosen as the baseline waste form by most of the countries to immobilize these waste streams, molybdate oxyanions (MoO 4 2- ) exhibit very low solubility (~1 mol %) in these glass matrices. In the past three to four decades, several studies describing the compositional and structural dependence of molybdate anions in borosilicate and aluminoborosilicate glasses have been reported in the literature, providing a basis for our understanding of fundamental science that governs the solubility and retention of these species in the nuclear waste glasses. However, there are still several open questions that need to be answered to gain an in-depth understanding of the mechanisms that control the solubility and retention of these oxyanions in glassy waste forms. This article is focused on finding answers to two such questions: (1) What are the solubility and retention limits of MoO 3 in aluminoborosilicate glasses as a function of chemical composition? (2) Why is there a considerable increase in the solubility of MoO 3 with incorporation of rare-earth oxides (for example, Nd 2 O 3 ) in aluminoborosilicate glasses? Accordingly, three different series of aluminoborosilicate glasses (compositional complexity being added in a tiered approach) with varying MoO 3 concentrations have been synthesized and characterized for their ability to accommodate molybdate ions in their structure (solubility) and as a glass-ceramic (retention). The contradictory viewpoints (between different research groups) pertaining to the impact of rare-earth cations on the structure of aluminoborosilicate glasses are discussed, and their implications on the solubility of MoO 3 in these glasses are evaluated. Lastly, a novel hypothesis explaining the mechanism governing the solubility of MoO 3 in rare-earth containing aluminoborosilicate glasses has been proposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Impact of High Performance Computing on the Detailed Simulation of Subsurface Flow and Contaminant Transport Using eSTOMP - 16341

The ability to model subsurface processes at the field-scale is essential to site closure and remedial action decisions. Field-scale models often require high-performance computing to meet the computational demands associated with complex systems, model calibration, and uncertainty quantification. eSTOMP, a parallel version of STOMP, is an analytical tool that meets the need for investigating coupled processes in the subsurface involving regional multi-fluid flow and biogeochemistry. The scalability of eSTOMP makes it ideal for simulating contaminant and geochemical reactive transport when long run times result from execution of a serial code. To demonstrate the impact of high-performance computation on subsurface analyses, the application of eSTOMP is presented for two sites. The first site involves uranium surface complexation within a highly dynamic, variably saturated flow field. The use of this model in conjunction with continued field monitoring is expected to provide a rigorous basis for developing operational strategies for field remediation and for defining defensible remediation endpoints. The second site involves the long-term dissolution of low-level radioactive glass emplaced within a near-surface repository. The waste form release rate is evaluated by modeling the basic physical and chemical processes that are known to control the waste form dissolution behavior, including pH and composition of the fluid contacting the glass. Simulations demonstrate that glass dissolution rates vary both in time and as a function of position within the disposal system, and that a 'leach rate' or radionuclide release rate parameter can only be estimated after identifying these impacts in a dynamic system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Iodobutane Deep Bed Adsorption Test Report

The Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has supported research and development on iodine control and iodine waste forms for the past several years. As a part of this on-going research, two deep-bed iodobutane adsorption tests have been performed with higher NO x levels. High (>1,000) iodine decontamination factors (DFs, the ratios of the inlet iodine concentrations divided by the outlet iodine concentrations, when the gas flowrate is constant) were achieved with both the silver aerogel and silver zeolite sorbents. The iodine adsorption mass transfer zone (the depth of the sorbent bed in which adsorption is occurring at any specific time) for these tests was estimated at 4 inches for both sorbents under the test conditions. Silver utilization for silver aerogel sorbent reached 41% (up to 12 g adsorbed iodine per 100 g silver aerogel sorbent), while silver utilization reached 43% (up to 6.8 g adsorbed iodine per 100 g silver zeolite sorbent). These values are generally consistent with results for prior methyl iodide adsorption tests. The results also indicate that, during iodine adsorption process, the organic iodide molecule decomposes so that the iodine can be adsorbed. Even when sorbent breakthrough is reached, the organic iodide molecule continues to decompose; and the unadsorbed iodine that passes through bed is no longer in the form of the original organic iodide; but in the form of iodine compounds that are soluble in NaOH solutions, such as HI or I 2 . This research and development has included iodine adsorption tests using a laboratory-scale test system containing segmented fixed beds of iodine adsorbents in series. Non-radioactive synthetic gas mixtures were blended from air, nitrogen, NO, NO 2 , water, and the target iodine compounds to represent dissolver off-gas (DOG) streams from used nuclear fuel aqueous reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Solidification of REDC Organics and Airborne Release Fraction Measurements

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 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 debris waste. The documented safety analysis for REDC uses an airborne release fraction of 5 x 10 -4 with a respirable fraction of 0.2 for plutonium and other actinides from packaged solid debris waste during a fire scenario, for an overall airborne release fraction/respirable fraction of 10 -4 . Solidifying the organic liquids, either by using PM-199 Organoclay ® or by grouting, would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. The use of Organoclay would be much simpler to implement inside a hot cell. 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. Grouting the liquids would require using multiple ingredients and mixing steps. Both methods produced waste forms that trapped cerium, as a surrogate for plutonium, inside the waste, with airborne release fractions during fire scenario tests of well below the 10 -4 limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Deep Borehole Disposal Research in the United States - 17427

The U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) manages a research and development (R and D) program for the disposal of spent nuclear fuel and high-level radioactive waste in the United States. Deep boreholes are being evaluated for disposal of some smaller DOE-managed waste as an alternative to disposal in a mined geologic repository. The deep borehole disposal concept consists of drilling borehole(s) to a total depth of about 5,000 m (16,400 ft - or greater than three miles) into crystalline basement rock, placing waste packages in the lower emplacement zone portion of the borehole, and sealing and plugging the upper portion of the borehole. The required bottom-hole diameter of the borehole(s) depends on the waste package configuration for the reference concept, but ranges from 22 to 43 cm (8.5 to 17 inches). Additional research and development is desired in several important areas for further consideration of deep borehole disposal of radioactive waste. A Deep Borehole Field Test (DBFT) will be conducted by DOE to further assess the viability of deep borehole disposal as an alternative for the disposal of smaller DOE-managed waste forms. The DBFT will be used to evaluate three main areas: 1) evaluation of the capability for drilling and construction of deep, large-diameter boreholes; 2) downhole scientific analyses to assess hydrogeochemical conditions that control waste stability and containment; and 3) engineering activities to assess the viability and safety of waste package emplacement and retrieval in deep boreholes. The DBFT will include first drilling a characterization borehole, followed by an optional field test borehole, to a total depth of about 5,000 m (16,400 feet) into crystalline basement rock in a geologically stable continental location. The initial smaller diameter characterization borehole will be drilled and completed to facilitate downhole scientific testing and analysis. If site conditions are found to be suitable, DOE intends to drill a larger-diameter field test borehole to evaluate emplacement activities using surrogate waste packages. The surrogate waste packages will not contain any radioactive waste, nor will any radioactive waste be used in these tests. One of the advantages of deep borehole disposal is that it would give DOE the flexibility to consider options for disposal of smaller waste forms. Deep borehole disposal also potentially offers earlier disposal of some wastes than might be possible in a mined repository, and could reduce costs associated with projected treatments of some wastes. In order to accomplish the DBFT, DOE will require the support of multiple national laboratories, along with a site contractor to acquire a suitable test site and to manage the site and the required testing activities. DOE issued a competitive solicitation via a request for proposal (RFP) in July, 2015, which requested proposals to provide a test site and site management services. The contract was awarded in January, 2016, to a team led by Battelle Memorial Institute which included a proposed test site in Pierce County, North Dakota. After efforts to acquire both the initial test site in North Dakota and an alternative proposed site in Spink County, South Dakota were unsuccessful, activities were suspended. The experiences in Pierce County, ND and Spink County, SD highlighted the importance of public engagement and support for the DBFT, and that relevant levels of government and other public stakeholders should be involved from the beginning. Using the lessons learned in North and South Dakota, DOE issued a new RFP on August 22, 2016 which emphasized local, state, and tribal (if applicable) government engagement, as well as public and other stakeholder involvement ahead of proposal submittals and throughout the contract execution phases. The new RFP allowed for multiple initial awards and multiple phases of contract execution. In December, 2016, DOE announced the selection of four companies - AECOM, ENERCON, RE/SPEC, and TerranearPMC - to begin exploring the possibility of conducting a DBFT under the terms of the new RFP. Ultimately, only one site will be chosen for the field test. (authors)

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

April 2025 Semiannual Composite Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from the Salt Waste Processing Facility is sampled semiannually for transfers to the Saltstone Production Facility. Salt solution is treated at the Saltstone Production Facility and disposed of in the Saltstone Disposal Facility. Per request of customer, X-TTR-Z-00027, Revision 0, one Saltstone Disposal Facility waste form was prepared at Savannah River National Laboratory using the Salt Waste Processing Facility Decontaminated Salt Solution and Z-area premix material for the April 2025 semiannual Toxicity Characteristic Leaching Procedure sample. The sample contained 60:40 (by weight) of slag and fly ash (referred to as the “Cement-Free grout sample”). Results from the technical report support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request prepared by Savannah River Mission Completion. At 43 days cured, the Saltstone sample was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure. The April 2025 semiannual Cement-Free grout sample met the South Carolina Code of Regulations for Hazardous Waste Management Regulations 61-79.261.24 and 61-79.268.48 requirements for a non-hazardous waste form with respect to the Resource Conservation and Recovery Act metals and Underlying Hazardous Constituents, and met the Saltstone Production Facility Waste Acceptance Criteria that was in effect at the time of the sample collection at the Salt Waste Processing Facility.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Thermal treatment of plutonium contaminated materials (PCM) Waste - 15714

Considering the high over all costs of waste disposal and the growing requirements for improved quality of the final waste form, the benefits offered by thermal processing become very significant. Key drivers for the application of thermal treatment processes include the reduced volume, improved passive safety, and superior long term stability, of the vitrified waste form products. The project has demonstrated these benefits are derived from the oxidation of the metallic waste fraction, destruction of organic components, and evaporation of entrained water, combined with simultaneous immobilization of radioactive and chemo-toxic elements with in a glass or slag (i.e. a partially crystallised glass) material. Currently a fundamental lack of scientific knowledge and understanding significantly hinders the uptake of thermal treatment processes for the immobilization and disposal of plutonium contaminated material waste. The project will contribute to accelerating the acquisition of knowledge and experience required to support the NDA in deploying thermal technologies as a national asset for ILW treatment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗