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At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

HPGe detector shielding optimization with MCNPX for the MEDINA PGNAA cell

Radioactive waste repositories must guarantee the non-toxicity of the waste in the long term, not only regarding radioactivity but also regarding other environmental contamination such as toxic chemicals. Analytical methods already exist for chemical characterization (ICP-MS, ICP-AES...) but they are based on test sampling. A possible alternative, for waste packages with an appropriate gamma radiation level, is to use Prompt Gamma Neutron Activation Analysis (PGNAA), a non-destructive measurement technique sensitive to several toxic chemicals. In view of the characterization of radioactive wastes in Germany and France, collaboration between the CEA Cadarache (France) and the Forschungszentrum Juelich (Germany) was initiated a few years ago. FZJ holds a PGNAA graphite cell called MEDINA (Multi Element Detection based on Instrumental Neutron Activation), allowing the characterization of 225 L drums. Fast neutrons are emitted from a D-T pulsed 14 MeV neutron generator and thermalized in graphite to induced radiative captures in the waste materials. Prompt capture gamma rays are detected using a 104% relative efficiency n-type HPGe. However, HPGe crystal is sensitive to fast neutron damage and to thermal neutron activation. A thermal neutron shield made of lithium fluorine and lithium carbonate is already used around the detector. In order to further decrease the current of fast and thermal neutrons coming into the crystal without penalizing MEDINA sensitivity (by decreasing the thermal neutron flux and neutron die away time of the cell, the gamma detection efficiency, or increasing the gamma background), some configurations based on easy-to-implement modifications of MEDINA have been simulated with MCNPX with a model of the cell already validated by experiments. Results show that fast and thermal neutron incoming current in the HPGe could easily be reduced by about a factor of 2 by additional quantities of graphite and by replacing lithium carbonate by lithium fluorine with a higher {sup 6}Li concentration. In addition, these modifications slightly increase the thermal neutron flux in the cell without deteriorating the neutron die away time, and reduce the gamma background about a factor of 2 during the neutron pulse but 5 times less after it. More important changes have also been tested, such as the addition of polyethylene and lead between the neutron generator and the HPGe detector, which is more effective regarding neutron shielding but decreases the neutron die away time, partly compensated by a larger initial thermal neutron flux. Concerning gamma background, hydrogen capture gamma ray (2.23 MeV) is increased due to the presence of polyethylene but lead around the HPGe decreases the total gamma background. In conclusion, simple modifications are possible to improve detector shielding and life time before thermal annealing of the crystal, without reducing MEDINA cell performances. Some of these modifications will be tested in the coming months. (authors)

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DOE OSTI2019

Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

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DOE OSTI2025

Annual Summary Report for the Remote-Handled Low-Level Waste Disposal Facility—FY 2024

This Fiscal Year (FY) 2024 annual summary report (ASR) documents the continued adequacy of the performance assessment (PA), the composite analysis (CA), and associated operating disposal- authorization statement (ODAS) technical-basis documents for the Remote-Handled (RH) Low-Level Waste (LLW) Disposal Facility at Idaho National Laboratory (INL). Annual review of the adequacy of the PA and CA for the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of Department of Energy (DOE) Order 435.1, “Radioactive Waste Management.” In FY 2024, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility-closure design, or institutional controls. Nineteen waste canister shipments were received at the RHLLW Disposal Facility, and all nineteen waste canisters were emplaced in disposal vaults.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2016

A Degradation Model and Testing Protocol for Metallic Waste Forms

Metallic waste forms developed to immobilize high-level radioactive wastes generated in advanced (closed) fuel cycles are being evaluated as part of the US DOE Fuel Cycle Research and Development program. The electrometallurgical reprocessing of used metallic fuels to recover U and Pu generates two waste streams: salt waste comprised of electro-refiner salt contaminated with oxidized fission products and metallic waste comprised of activated cladding and metallic fission products that were not oxidized under the processing conditions. Metallic waste forms can be made by directly melting the cladding and fuel wastes with small amounts of trim metals to form a multiphase alloy material in which radionuclides and other waste constituents are immobilized within chemically durable intermetallic and iron-based solid solution phases. Some components of the waste streams serve as important constituents of phases comprising the waste form, such as Mo and Zr, which form Fe{sub 2}Mo and Fe{sub 2}Zr intermetallic phases. Other waste constituents are accommodated as substitutional contaminants, such as Tc within the Fe{sub 2}Mo intermetallic and U/TRU within the Fe{sub 2}Zr intermetallic. Sufficient amounts of these host phases must be available to accommodate all radionuclides in the waste stream. The compositional flexibility of the host phases to incorporate waste components is being utilized to formulate metal waste forms that will accommodate variations in waste streams compositions that will occur over long-term fuel reprocessing operations. Small amounts of trim metals can be added to tailor the phase assembly to match the waste stream composition and facilitate waste form processing. For example, additional Zr may be added during waste form production to ensure all U/TRU is sequestered in a durable Fe{sub 2}Zr host phase and trim Cr, Mo, and Ni can be added to supplement the amounts provided by the steel cladding to improve the durability of a metallic waste form. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Electrical Resistivity Changes During Heating Experiments Unravel Heterogeneous Thermal‐Hydrological‐Mechanical Processes in Salt Formations

Abstract Rock salt is considered a suitable medium for the permanent disposal of heat‐generating radioactive waste due to its isolation properties. However, excavation damage and heating induce complex and heterogeneous thermal‐hydrological‐mechanical (THM) processes across different zones. Quantifying this heterogeneity is crucial for accurate long‐term performance assessment models, but traditional methods lack the necessary resolution. This study employs 4D electrical resistivity tomography (ERT) monitoring during controlled heating experiments in a salt formation to unravel the spatiotemporal dynamics of THM processes. Advanced time‐lapse inversion and clustering analysis quantify subsurface properties and map the heterogeneity of THM dynamics. The ERT results can estimate subsurface properties and delineate the damaged and intact zones, enabling appropriate parameterization and representation of processes for long‐term modeling. This approach may be used in further improving the predictive models and ensuring the safe long‐term disposal of radioactive waste in rock salt.

58 GEOSCIENCES↗
DOE OSTI2016

Coupling compositional liquid gas Darcy and free gas flows at porous and free-flow domains interface

This paper proposes an efficient splitting algorithm to solve coupled liquid gas Darcy and free gas flows at the interface between a porous medium and a free-flow domain. This model is compared to the reduced model introduced in [6] using a 1D approximation of the gas free flow. For that purpose, the gas molar fraction diffusive flux at the interface in the free-flow domain is approximated by a two point flux approximation based on a low-frequency diagonal approximation of a Steklov–Poincaré type operator. The splitting algorithm and the reduced model are applied in particular to the modelling of the mass exchanges at the interface between the storage and the ventilation galleries in radioactive waste deposits.

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DOE OSTI2018

Studies on Workability of Geopolymer Solidification - 18171

In Japan, intermediate-level radioactive waste, such as spent resin, is solidified and set in sealed containers without venting. To reduce the amount of hydrogen generated due to water radiolysis in stored intermediate-level waste, geopolymer solidification has been studied. Geopolymer solidification is an excellent method for conditioning intermediate-level waste from the viewpoint of its high thermal stability and low processing cost. However, it has a problem that the apparent viscosity increases drastically a few minutes after the onset of setting. This may cause some trouble in the process of pouring the geopolymer into drums and mixing it with waste. In this study, in order to control the workability of a geopolymer mixture, we investigated the effects of the compounding ratio and the kinds of chemical agents used. We confirmed that the geopolymer reaction was slowed down by reducing the amount of calcium and adding boron. The uniaxial compressive strength of the geopolymer was 3-80 MPa. (authors)

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DOE OSTI2018

The 'Site of Things': Remediation in the Information Age - 18088

More than just a payload, machines now haul data as well. Wood Environment and Infrastructure's maiden voyage into civil engineering automation at the Port Granby Project Long-Term Waste Management Facility demonstrated enormous potential for the remediation industry. The Port Granby Project will relocate approximately 450,000 cubic meters of historic low-level radioactive waste and marginally contaminated soils, located at an existing waste management facility on the shoreline of Lake Ontario, to a new, engineered aboveground mound to be built about a kilometer north of the current site. Sounds simple, right? Just move a bunch of dirt up the road? Hardly, this is marginally contaminated soil situated near Lake Ontario in a beautiful little village an hour east of Toronto. We must relocate safely, meaning limiting the number of staff in the field and efficiently to deliver according to our agreed upon budget and schedule. As if that wasn't enough we were asked to keep track of where each truckload of dirt was excavated from (with a high degree of accuracy), where it was placed in the new engineered disposal facility, how much it weighed and the amount of radioactivity in each load... a virtual chain of custody for each load. Presented is an overview of the Information Technology that supports the Port Granby Project and how the Data Management System (DMS) is being leveraged to make the project safer, more effective and efficient. (authors)

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DOE OSTI2018

French Innovative Processes in the Field of Thermal Treatment For Decommissioning and Legacy Waste - 18085

In order to optimize decommissioning waste management and better anticipate related issues, French nuclear waste producers CEA and AREVA, and the French Agency in charge of radioactive waste disposal (ANDRA) have decided to collectively develop technologies mastering costs, schedules and dose uptake as well as optimizing waste volumes and repository safety. In this context, several innovative projects have emerged over the past few years to develop specific thermal treatment processes. This paper focus on four new processes, presenting their goals, technical descriptions and first experimental results. (authors)

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DOE OSTI2017

Knowledge Management in the Field of Nuclear Energy - How to Bridge the Gap between Retiring Workforce and Next Generations - 17101

The current phase-out decision of the German Federal Government and the 'Energiewende' - resulting from the Fukushima Daiichi Accident in 2011 - changed the nuclear market significantly. Together with the change in political and social attitude towards nuclear energy since 2002 it has contributed to a decline in technical expertise. This as well as the generation change in the nuclear technology lead to the risk of losing accumulated knowledge and experience which still are essential for the future - not only in Germany but in other countries - even though the focus in Germany will shift from operating to decommissioning, dismantling and radioactive waste management. A long-term strategy in order to preserve technical expertise to solve the forthcoming tasks in Germany was required. This strategy should include e.g. the implementation of technical programs not only for internal training of the staff but also for authorities. Together with authorities TUV NORD developed individual training concepts for a broad range of customers to preserve a transfer of know-how with the state-of-the-art in science and technology and even support international projects, especially in countries with less experience in the nuclear fields. Additionally, special data bases were developed in order to ensure the accessibility of long-term nuclear data storage. Furthermore technical conferences in different fields like decommissioning and radioactive waste disposal were established as an opportunity for knowledge exchange. Based on given examples it is being demonstrated how technical expertise can be preserved even though the field of nuclear energy in Germany has changed. (authors)

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DOE OSTI2016

Deep Borehole Field Test Laboratory and Borehole Testing Strategy

Deep Borehole Disposal (DBD) of high-level radioactive wastes has been considered an option for geological isolation for many years (Hess et al. 1957). Recent advances in drilling technology have decreased costs and increased reliability for large-diameter (i.e., ≥50 cm [19.7”]) boreholes to depths of several kilometers (Beswick 2008; Beswick et al. 2014). These advances have therefore also increased the feasibility of the DBD concept (Brady et al. 2009; Cornwall 2015), and the current field test design will demonstrate the DBD concept and these advances. The US Department of Energy (DOE) Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013) specifically recommended developing a research and development plan for DBD. DOE sought input or expression of interest from States, local communities, individuals, private groups, academia, or any other stakeholders willing to host a Deep Borehole Field Test (DBFT). The DBFT includes drilling two boreholes nominally 200m [656’] apart to approximately 5 km [16,400’] total depth, in a region where crystalline basement is expected to begin at less than 2 km depth [6,560’]. The characterization borehole (CB) is the smaller-diameter borehole (i.e., 21.6 cm [8.5”] diameter at total depth), and will be drilled first. The geologic, hydrogeologic, geochemical, geomechanical and thermal testing will take place in the CB. The field test borehole (FTB) is the larger-diameter borehole (i.e., 43.2 cm [17”] diameter at total depth). Surface handling and borehole emplacement of test package will be demonstrated using the FTB to evaluate engineering feasibility and safety of disposal operations (SNL 2016).

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DOE OSTI2015

The evolution of the Nevada National Security site waste acceptance criteria - 15312

The Nevada National Security Site Waste Acceptance Criteria (DOE/NV--325) is an evolution of requirements spanning almost three decades of waste acceptance at the Nevada National Security Site (NNSS). The original waste acceptance criteria (WAC) document was written in 1978 (document NVO-185) and underwent multiple revisions as time passed and disposal experience was gained. This document came into being when the NNSS first started receiving significant quantities of low-level radioactive waste (LLW) from offsite U.S. Department of Energy (DOE) generators, such as the Mound and Rocky Flats Plants. The original WAC was far less prescriptive and rigorous than the current criteria since many of the current DOE order directives, Federal/State laws, and regulations did not exist. The current WAC reflects the results of numerous revisions in order to address the nuances of the varied waste streams and package configurations accepted for disposal at the NNSS. This includes waste streams that exhibit challenges not previously faced or covered by the WAC and therefore, from time to time, require new information and or changes to be made to the WAC. These wastes now represent a growing percentage of the remaining waste at cleanup sites across the DOE Complex. These waste streams present challenges, such as the technical basis for ensuring safe transportation to and disposal at the NNSS. The current WAC (Rev. 10 dated June 2013), is Available online at: http://www.nv.energy.gov/library/publications/Environmental/DOENV{sub 3}25.pdf). Other challenges for the NNSS include incorporating new and emerging technologies and addressing stakeholder concerns. Addressing these challenges sometimes results in changes to the NNSS WAC. Drivers for revising the NNSS WAC over the years have included operating experience improvements, the need to specify waste origin/ownership, and identifying preferred transportation routing. This paper will discuss various aspects of the NNSS WAC, and how it has evolved based upon DOE orders and requirements, acceptance of offsite mixed wastes, acceptance of non-radioactive classified wastes, stakeholder concerns and agreements, and the changing face of the remaining legacy waste at DOE sites across the United States. (authors)

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DOE OSTI2025

DECOVALEX-2023: An international collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems

The DECOVALEX initiative is an international research collaboration (www.decovalex.org), initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. DECOVALEX stands for “DEvelopment of COupled Models and VALidation against EXperiments”. The creation of this international initiative was motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. DECOVALEX emphasizes joint analysis and comparative modeling of the complex perturbations and coupled processes in geologic repositories and how these impact long-term performance predictions. The most recent phase of the DECOVALEX Project, here referred to as DECOVALEX-2023, started in early 2020 and ended in late 2023. More than fifty research teams associated with 17 international DECOVALEX partner organizations participated in the comparative evaluation of eight modeling tasks covering a wide range of spatial and temporal scales, geological formations, and coupled processes. This Virtual Special Issue on DECOVALEX-2023 provides an in-depth overview of these collaborative research efforts and how these have advanced the state-of-the-art of understanding and modeling coupled THMC processes. While primarily focused on radioactive waste, much of the work included here has wider application to many geoengineering topics.

Coupled processes↗
DOE OSTI2023

Spectroscopic identification of Ca-bearing uranyl silicates formed in C–S–H systems

Portland cement-based grouts used for radioactive waste immobilisation contain a Ca- and Si-rich binder phase, known as calcium–silicate–hydrate (C–S–H). Depending on the blend of cement used, the Ca/Si ratio can vary considerably. A range of C–S–H minerals with Ca/Si ratios from 0.6 to 1.6 were synthesised and contacted with aqueous U(VI) at 0.5 mM and 10 mM concentrations. Solid-state 29 Si MAS-NMR spectroscopy was applied to probe the Si coordination environment in U(VI)-contacted C–S–H minerals and, in conjunction with U L III -edge X-ray absorption spectroscopy analysis, inferences of the fate of U(VI) in these systems were made. At moderate or high Ca/Si ratios, uranophane-type uranyl silicates or Ca-uranates dominated, while at the lowest Ca/Si ratios, the formation of a Ca-bearing uranyl silicate mineral, similar to haiweeite (Ca[(UO 2 ) 2 Si 5 O 12 (OH) 2 ]·3H 2 O) or Ca-bearing weeksite (Ca 2 (UO 2 ) 2 Si 6 O 15 ·10H 2 O) was identified. This study highlights the influence of Ca/Si ratio on uranyl sequestration, of interest in the development of post-closure safety models for U-bearing radioactive waste disposal.

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DOE OSTI2019

Remediation of Irradiated Graphite Using Microorganisms Isolated from Oil Contaminated Soil - 19643

The deployment of nuclear energy globally is hindered by the fear cleared by nuclear reactor disasters such as the Chernobyl and Fukushima reactor meltdowns. Additionally, accumulation of radioactive waste globally will, in the near future, present a significant threat to humans and the environment. Notably, the majority of nuclear waste from the energy generation industry is generated as Low Level Waste (LLW) in which the small radioactive component renders the bulk of the waste unsuitable for conventional treatment and disposal. In the case of radioactive graphite, the small radioactive component comprising radiocarbon-14 (C-14) and metallic pollutants can be removed to allow safe disposal or reuse as of the graphite. In the current study, mixed-cultures of organisms isolated from oil contaminated sites showed potential of selective bioaccumulation of C-14 from a powdered C-12/C-14 graphite matrix. The study was conducted in a continuously mixed reactor (growth vessel) constructed from Pyrex panels (30 cm x 20 cm x 15 cm) with finely ground graphite as the principal carbon source (1500-2000 mg.L{sup -1}) and spikes of glucose (500 mg.L{sup -1} every 48 hours). The growth vessel was connected to a biofilter unit by submersible pump where particulate matter was isolated for further analysis. Results showed that approximately 68% more C-14 was released from ignited dry cells than from respiration product. C-14 accumulated in cells corrected using a trap filter system, with a marked decrease of C-14 in carbon dioxide than vegetative suspended cells in the main tank. The only mechanism that could account for the differential C-14 distribution between the liquid medium and cells was possible fractionation during uptake of carbon compounds. The discriminating factor suggested in this study is the kinetic isotope effect (KIE) which allows heavier atoms to have different affinities for reaction sites on catalysis and biocatalysis. (authors)

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DOE OSTI2017

Waste Management Strategies for Production of Mo-99

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.

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