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Results for “RADIONUCLIDE MIGRATION”

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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Influences of temperature on permeability changes of flow-paths altered by highly alkaline Ca-rich groundwater - 15192

Fractures (or cracks) in host rock around the geological repository are considered as main flow-paths for the migration of radionuclides. Such fractures would be altered by the alkaline components (Ca, Na, K) leaching from cementitious materials for the construction of the repository. While silicate minerals in fractures are dissolved by highly alkaline groundwater, secondary minerals would deposit on the surface of fractures because of the reaction between the alkaline components and silicate species. The authors' previous study showed the generation of CSH-gel and the clogging of micro flow-path under the condition of Ca-rich and high pH by flow experiments. The present study examined the permeability changes with temperature in micro flow-path flowing highly alkaline Ca-rich solution, considering temperature at the depth of geological disposal system. For the simulation of micro flow-path such as fractures, 'micro flow-cell', which nipped a thin Teflon sheet with a slit between granite chip and Teflon basement, were used for the experiments. Highly alkaline Ca-rich solution (pH>12 and 8.5 mM Ca(OH){sub 2}) were injected into micro flow-cell with 2.0 ml/h of flow rate under the given temperature conditions (278 K, 296 K or 313 K). As a result, the permeability became lower with the increase in temperature. In addition, the deposition rates of CSH-gel became larger with the increase in temperature, and the activation energy was estimated to be 32.0 kJ/mol. These results suggest that flow-paths around the geological repository would undergo the clogging with CSH-gel in a shorter time-period under a condition of higher temperature. (authors)

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

Portland cement paste behaviour under gamma radiation field in geological repository - 15272

The Radioactive Waste Management Laboratory at the Nuclear and Energy Research Institute in Sao Paulo, Brazil (IPEN-CNEN-SP), is developing the concept of a repository for disposition of disused sealed radioactive sources in a deep borehole in which cement paste is intended to be used as a backfill between the steel casing and the geological formation around the borehole. The cement paste is intended to function as structural material, a blockage against the transport of water between the different strata of the geological setting, and as an additional barrier against the migration of radionuclides. The assessment of the cement paste behaviour in long term is necessary to increase the confidence that the material will perform as required during the service life of the facility. Under radiation, pore water present in cement paste suffers radiolysis and some products can be highly reactive, as electrons and hydroxyl radicals and hydrogen peroxide. These radiolysis products will interact with cement paste and its hydration products, forming a wide range of compounds. Previous work focused on assessing the synergic effects of radiation and environmental conditions, like higher temperature and aggressive groundwater chemicals, on cement paste. Examination of samples detected small differences in mineralogy and microstructure. However, the radiation dose to which samples were exposed (400 kGy) was only a fraction of the expected accumulated dose in cement paste until complete decay of some sealed sources in the repository. Therefore, several thousand kilo-grays must be tested the possibility of radiation to alter cement mineralogy and, as a consequence, its durability The objective of this research is to evaluate the changes in mineralogy and microstructure of irradiated cement paste specimens, in an attempt to establish a relationship between irradiation and durability. Cubic Portland cement specimens were casted and exposed to the gamma radiation of a 60 Co multipurpose irradiator, at IPEN-CNEN/SP. The effects of doses between 1000 kGy and 4000 kGy are evaluated by X-Ray Diffraction, Scanning Electron Microscopy, X-ray micro tomography, neutrongraphy and neutron scattering. This report presents the changes in cement paste mineralogy when samples are examined by XRD. Results indicate that durability can be affected because of the observed changes. Results are expected to indicate the induced changes in cement paste under high radiation doses and to provide hints on its behaviour in the long-term and service-life. The results can be used in safety analysis for deep geological repositories that use cement based materials as engineered barriers. (authors)

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

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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

Effects of pH on deposition rate of supersaturated silicic acid around geological disposal system - 15245

For the migration assessment of radionuclides around geological repository, it is important to estimate the dynamic behavior of silicic acid under a high alkalinity condition due to the cementitious materials for the construction of the repository. This study examined the pH dependencies of the deposition behavior of supersaturated silicic acid in the co-presence of solid phase. The parameters for the deposition experiments were as follow: the pH was set to 8, 9 and 10, the initial saturated concentration was set to 2-8 mM, the amount of amorphous silica (350 m{sup 2}/g) as solid phase was set to 0.1-1.0 g. The experimental results showed that the initial supersaturated concentrations strongly affected the formation of colloidal silicic acid, increasing the deposition amount of silicic acid and decreasing the soluble silicic acid with time. In other words, the polymerization and the deposition of silicic acid simultaneously arose in these experiments. Furthermore, this study evaluated the apparent deposition rates of silicic acid by assuming the first-order reaction equation considering the specific surface area. As a result, the deposition rate constants, k, were in a narrow range of 10{sup -11} to 10{sup -10} m/s under all experimental conditions in this study. This means that the deposition rate constants do not depend both on pH-value (in the range from 8 to 10) and the initial concentration of supersaturated silicic acid. That is, the estimated deposition rate constants are useful in estimation of the deposition behavior of supersaturated silicic acid, such as the clogging in fractures (flow-paths) around the repository, even if the pH-value of groundwater gradually or dramatically changes. (authors)

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

On the long-term migration of uranyl in bentonite barrier for high-level radioactive waste repositories: The effect of different host rocks

Repositories for high level radioactive waste hosted in argillite or granite with bentonite backfill are the two disposal concepts that been extensively studied. A general perception is that granite has very low sorption capacity and therefore is disadvantageous in retarding the migration of radionuclides in comparison with argillite. However, when there is bentonite backfill, does the sorption capacity of granite really matter for the migration of radionuclides? How different host rocks would affect the migration of U(VI) through the bentonite backfill of an engineered barrier system (EBS), and which properties of the host rocks cause those differences in migration, given that migration of radionuclides is a critical measure in assessing the performance of a repository? Here we present two coupled thermal, hydrological, and chemical models for the transport of U(VI), each with an identical setup except for the type of host rock (argillite vs granite rock). Comparisons between our models show that the different host rocks exert their influence on migration of U(VI) via regulating the chemical conditions in the bentonite, which affect the concentration of U(VI) at the source and the adsorption of U(VI) in the bentonite. The key chemical conditions are pH, Ca + , and HCO 3 - concentration in pore-water, which are strongly affected by soluble carbonate minerals. Our models also show the occurrence of illitization (dissolution of smectite and precipitation of illite) in the bentonite, which affects the migration of U(VI) through changing pH and the quantity of adsorbents. Although generalization of current model results for granite and argillite should be done carefully, the simulations highlight the importance of pore-water chemistry in host rock, as well as bentonite-host rock interactions, when assessing the performance of a repository. In comparison to properties such as thermal conductivity, permeability, and adsorption capacity, properties that have been studied intensively, the pore-water chemistry of host rocks deserve equal or even more attention because of its strong influence on radionuclide migration.

58 GEOSCIENCES↗
DOE OSTI2024

DECOVALEX-2023: Task F2 Salt Final Report

The subject of Task F of DECOVALEX-2023 concerns performance assessment modelling of radioactive waste disposal in deep mined repositories. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. In Task F2- (salt), these objectives have been accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Coupled-process submodels and deterministic simulations of the entire PA model for a reference scenario for waste disposal in domal salt have been conducted. The task specification has been updated continuously since the initiation of the project to reflect the staged development of the conceptual repository model and performance metrics. Thermal, hydrological, mechanical, and chemical properties of individual components of the engineered and natural system were chosen for relevance by participating teams. The salt reference case system was characterized using data and measurements collected at relevant underground research laboratories (URLs), field sites, and simulation results from teams with specialized modelling capability. Participating teams made a wide range of model assumptions from compartmentalized networks to full 3D models of the salt formation. No single contributed model includes full-fidelity representation of all the features, events, and processes (FEPs) detailed in the task specification, but almost all features and processes are represented in at least one model. Despite differences in the modelling strategies developed by participating teams, all models indicate that salt compaction and radionuclide diffusion are key processes in the repository, and for the FEPs and model scenario considered, little of the disposed radionuclides will migrate beyond the repository seal over the 100,000 year simulations. In general, the model output quantities have the largest differences over the short term and near the waste. The models tend to be more similar further from waste and at later time. Disparities between the models are believed to be due to differing simplifications from the task specification, some of which are chosen simplifications to reduce complexity, and some are restrictions imposed by the modelling tools. A second round of this task has been accepted for DECOVALEX-2027 in conjunction with Task F1 on crystalline PA modelling. The future round includes waste package heating, improved modelling of salt creep closure, additional comparisons of coupled-process sub-models, and the impact of repository engineering design on radionuclide migration in the repository. Participants will also propose and finalize a set of uncertain inputs for the reference case simulations, propagate these uncertainties in a set of realizations, and conduct sensitivity analyses on the simulation results.

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

Estimation of Cs Migration under an Unsaturated Condition by Parallel-plate Experiments and 2D Advection-dispersion Model - 17188

For the assessment of the radioactive waste facility in the near-surface underground, it is important to consider the influence of the unsaturated condition flow field (hereinafter referred to as unsaturated condition) on the migration of radionuclides through the porous medium. The unsaturated condition, which is not completely saturated with liquid phase, will change the retardation effect for the migration of nuclides by interrupting the interaction of nuclides and solid phase, and by clogging flow paths with gas phase. In this study, considering the spatial distribution of the water saturation, the migration of radionuclides in the two-dimensional (2D) porous medium under the saturated flow field (hereinafter referred to as saturated condition) and unsaturated condition was examined by using the parallel-plate type's packed bed and the 2D advection-dispersion model. Besides, for comparison, the one-dimensional (1D) flow experiment using the column type's packed bed was also examined. In the experiments, a 1 mM Cs solution was injected into the beds packed with silica sand under a given constant pressure condition controlled by an overflow system. The overall effective values of retardation coefficient were estimated by applying the theoretical models to the experimental results. As a result, the overall retardation coefficient under the unsaturated condition became smaller than that under the saturated condition. This means the decrease in the effective surface area for the sorption of tracer ions in the presence of gas phase. Furthermore, under the unsaturated condition, the retardation coefficient of the 2D flow condition was estimated smaller than that of the 1D condition. That is, in the 2D flow system, the tracer ions migrate through more selective flow-paths caused by the heterogeneous spatial-distribution of water saturation which is the ratio of the volume of water to the volume of pores. (authors)

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

Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

Assessments of waste form and disposal options start with the degradation of the waste forms and consequent mobilization of radionuclides. Long-term static tests, single-pass flow-through tests, and the pressurized unsaturated flow test are often employed to study the durability of potential waste forms and to help create models that predict their durability throughout the lifespan of the disposal site. These tests involve the corrosion of the material in the presence of various leachants, with different experimental designs yielding desired information about the behavior of the material. Though these tests have proved instrumental in elucidating various mechanisms responsible for material corrosion, the chemical environment to which the material is subject is often not representative of a potential radioactive waste repository where factors such as pH and leachant composition will be controlled by the near-field environment. Near-field materials include, but are not limited to, the original engineered barriers, their resulting corrosion products, backfill materials, and the natural host rock. For an accurate performance assessment of a nuclear waste repository, realistic waste corrosion experimental data ought to be modeled to allow for a better understanding of waste form corrosion mechanisms and the effect of immediate geochemical environment on these mechanisms. Additionally, the migration of radionuclides in the resulting chemical environment during and after waste form corrosion must be quantified and mechanisms responsible for migrations understood. The goal of this research was to understand the mechanisms responsible for waste form corrosion in the presence of relevant repository sediments to allow for accurate radionuclide migration quantifications. The rationale for this work is that a better understanding of waste form corrosion in relevant systems will enable increased reliance on waste form performance in repository environments and potentially decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

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

A Preliminary Performance Assessment for Salt Disposal of High-Level Nuclear Waste - 12173

A salt repository is one of the four geologic media currently under study by the U.S. DOE Office of Nuclear Energy to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic salt repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a salt formation. The current phase of this study considers representative geologic settings and features adopted from previous studies for salt repository sites. For the reference scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small. For the human intrusion (or disturbed) scenario, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario. Actinides including Pu-239, Pu-242 and Np-237 are major annual dose contributors, and the calculated peak mean annual dose is acceptably low. A performance assessment model for a generic salt repository has been developed incorporating, where applicable, representative geologic settings and features adopted from literature data for salt repository sites. The conceptual model and scenario for radionuclide release and transport from a salt repository were developed utilizing literature data. The salt GDS model was developed in a probabilistic analysis framework. The preliminary performance analysis for demonstration of model capability is for an isothermal condition at the ambient temperature for the near field. The capability demonstration emphasizes key attributes of a salt repository that are potentially important to the long-term safe disposal of UNF and HLW. The analysis presents and discusses the results showing repository responses to different radionuclide release scenarios (undisturbed and human intrusion). For the reference (or nominal or undisturbed) scenario, the brine flow rates in the repository and underlying interbeds are very low, and transport of radionuclides in the transport pathways is dominated by diffusion and greatly retarded by sorption on the interbed filling materials. I-129 (non-sorbing and unlimited solubility with a very long half-life) is the dominant annual dose contributor at the hypothetical accessible environment, but the calculated mean annual dose is negligibly small that there is no meaningful consequence for the repository performance. For the human intrusion (or disturbed) scenario analysis, the mean mass release rate and mean annual dose histories are very different from those for the reference scenario analysis. Compared to the reference scenario, the relative annual dose contributions by soluble, non-sorbing fission products, particularly I-129, are much lower than by actinides including Pu-239, Pu-242 and Np-237. The lower relative mean annual dose contributions by the fission product radionuclides are due to their lower total inventory available for release (i.e., up to five affected waste packages), and the higher mean annual doses by the actinides are the outcome of the direct release of the radionuclides into the overlying aquifer having high water flow rates, thereby resulting in an early arrival of higher concentrations of the radionuclides at the biosphere drinking water well prior to their significant decay. The salt GDS model analysis has also identified the following future recommendations and/or knowledge gaps to improve and enhance the confidence of the future repository performance analysis. - Repository thermal loading by UNF and HLW, and the effect on the engineered barrier and near-field performance. - Closure and consolidation of salt rocks by creep deformation under the influence of thermal perturbation, and the effect on the engineered barrier and near-field performance. - Brine migration and radionuclide transport under the influence of thermal perturbation in generic salt repository environment, and the effect on the engineered barrier and near-field performance and far-field performance. - Near-field geochemistry and radionuclide mobility in generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Degradation of engineer barrier components (waste package, waste canister, waste forms, etc.) in a generic salt repository environment (high ionic strength brines, elevated temperatures and chemically reducing condition). - Waste stream types and inventory estimates, particularly for reprocessing high-level waste. (authors)

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

Alteration Behaviors of Granite Minerals Forming Calcium Silicate Hydrate under Ca-Rich Highly Alkaline Conditions - 16160

Cementitious materials used for geological disposal systems leach large amounts of alkaline ions (Ca{sup 2+}, Na{sup +}, K{sup +}) into groundwater, and alter the pH of groundwater by up to 13. The altered groundwater contacts the host rock and forms Calcium Silicate-Hydrate (CSH) through the interaction of Ca ions and the silicate species, which in turn clogs the flow paths of the host rock. This clogging effect in the flow paths is expected to retard the migration of radionuclides. A previous study reported the heterogeneous formation of CSH on the surface of granite under Carich and highly alkaline conditions. To clarify the formation process of CSH, this study examined the CSH formation with the dissolution of some minerals contained in the granite under Ca-rich and highly alkaline conditions. In the experiments, the granite block and the crushed minerals contained in the granite were immersed in a 6.4-8.5 mM Ca(OH){sub 2} solution adjusted pH to 12.5 by adding NaOH solution. As a result, biotite and anorthite supplied more Si to the formation of CSH than other minerals. Furthermore, most of the formed CSH precipitated on the solid surface although a little CSH remained in liquid phase as supernatant. These results suggest that under Ca-rich highly alkaline conditions the CSH formed as a secondary mineral will retard the migration of radionuclides released from the repository because of the clogging effect of the flow path. (authors)

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

Sorption Behavior of Eu(III) into CSH Gel in Imitated Saline Groundwater - 12145

The sorption behavior of Eu(III) (europium (III)) into CSH (Calcium Silicate Hydrate) gel without dried processes was examined in imitated saline groundwater by using the spectro-fluorometer, Raman spectrophotometer and ICP-AES (Inductively Coupled Plasma- Atomic Emission Spectrometry). Ca/Si ratio was set to 0.4, 0.8, 1.2 and 1.6, and NaCl concentration was also set to 0.6, 0.06 and 0.006. The synthesis of each sample was conducted in a glove box saturated with nitrogen gas. The sealed sample tubes were gently shaken with 120 strokes/min. The time-period to contact Eu(III) with the CSH gel was set to 60 days. The fluorescence emission spectra suggested the incorporation of Eu{sup 3+} into CSH gel in the high Ca/Si ratio samples. On the other hand, from the decay behavior of fluorescence emission spectra, even in the low Ca/Si ratio samples, sorption behavior of Eu{sup 3+} into CSH gel was confirmed. Besides, the Raman spectra showed that the degree of polymerization of Si-O in CSH gel was raised with increasing Na ions concentration. These results suggest that the CSH gel, formed as secondary mineral, would retard the migration of radionuclides even in saline groundwater. Considering the inflow of saline groundwater into repository, this study examined the interaction between CSH gel (without dry processes) and Eu{sup 3+} by using the fluorescence emission spectra, the decay behavior of fluorescence and the Raman spectra. As a result, the fluorescence emission spectra of the sample of more than 0.8 Ca/Si ratio confirmed the intensity split into two peaks around 618 nm (5D0→7F2 transition). Furthermore, even in relatively low Ca/Si ratio samples, the fluorescence lifetimes both of the surface sorption sample and the co-precipitated samples exceeded that of the filtrate sample. These suggested that Eu{sup 3+} is not only hydrolyzed to form Eu(OH){sub 3} colloid, but is also stably incorporated into CSH gel (in Ca/Si ratio>1.2) or is forming complex on the surface of solid phase (in Ca/Si<0.8) in the co-presence of Na ions. On the other hand, the concentrations of Na, Ca and Si after 60 days in the solution filtrated through 0.2 μm membrane filter showed the ion exchange of Ca and Eu apparently in the samples of low Ca/Si ratio. Besides, in the samples with high Ca/Si ratio, Ca and Si concentrations in the solution were raised with increment of initial Na concentration. While Na ions may slightly increase the solubility of CSH gel, the CSH samples were stable, mostly maintaining the initial Ca/Si ratio synthesized. Moreover, the Raman spectra showed that the degree of polymerization of silicate chain in CSH samples increased with increasing Na concentration. Such an immobilized Na into the structure of CSH gel did not obstruct the incorporation of Eu{sup 3+} into CSH gel. The results mentioned above suggested that CSH gel (formed as a secondary mineral around the repository) also can retard the migration of radionuclides even if the repository and its surrounding are saturated by saline groundwater. (authors)

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

Percolation in Salt Rocks - 18652

We present laboratory investigations of fluid percolation in salt rocks, performed on drilled core samples, which show that natural rock salt does not become permeable at high pressures and temperatures, as had been claimed in the literature. Rather, we confirm the existence of a percolation threshold: Fluid migration is only possible when the fluid pressure equals or exceeds the lowest confining stress. The lab tests corroborate in-situ-determinations of the percolation threshold in salt formations: The threshold is given by the minimal stress in the rock mass (minimal stress criterion), and fluid pressure-driven percolation starts once it is exceeded. The hypothesis that the percolation threshold vanishes at high pressure and temperature is thus refuted by experiments on natural rock salt. These experimental tests, performed in the framework of the KOSINA project, are an important prerequisite for the proof of integrity of saliferous barriers in the disposal of radioactive waste in salt rocks. (authors)

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

A Decision-Based Iterative Modeling Framework for Probabilistic Performance Assessments - 18441

Radiological probabilistic performance assessments (PPAs) employ highly complex computer models to simulate the release of radionuclides from wastes in an engineered disposal facility, migration of these radionuclides over time, and resulting radiological dose to potentially exposed individuals. An iterative approach to PPA modeling is described to help define the appropriate complexity of the computer model. An initial model is developed that addresses all relevant processes and media, but that employs preliminary statistical parameter distributions to minimize the initial effort expended. Where appropriate, transport and exposure submodels with simplifying assumptions are also used in the initial model. Different assumptions and boundaries may be explored with the initial model to assess their importance in constraining model results. Exploration of the results of the initial model are used to determine the benefits of model refinements, and to guide such refinements to focus on specific areas of the model that most influence the model outputs of interest. This approach promotes efficiency and model parsimony, but does not compromise defensibility when performed in a transparent manner. Key model boundaries for a PPA include the time frame of the modeling (temporal boundaries), where future human receptors may be located (spatial boundaries), and what types of land uses and activities may exist in the future (exposure scenarios). Simplified transport and exposure models can be useful in helping decision makers and stakeholders establish such boundaries by analyzing the consequences of selecting different bounds and understanding which assumptions and boundaries have the greatest influence on the magnitude and distribution of model results. Pathway analysis and parameter sensitivity analysis performed on PPA model results are key components of the iterative modeling approach. The objective of pathway analysis is to identify the critical radionuclides and pathways contributing to radiation dose during the modeling performance period. An environmental 'pathway' is described by the connections from a waste package or cell in the disposal facility to a future human exposure event. These connections are commonly visualized using graphical means. A simple example of a potentially complete pathway for radon-222 is: radium-containing waste → radon emanation → upwards diffusion in cap material → diffusion/advection into a future building → inhalation exposure to a building occupant Parameter sensitivity analysis is conducted on the results of a PPA to identify those specific model input parameters whose distribution of values contributes the most to the distribution of target model output, such as annual radiation dose. Pathway and sensitivity analyses on initial modeling results provide information for guiding subsequent model development, and possibly interim decisions, by informing questions tied to the modeling objectives and performance criteria. For example, questions related to radiological doses from existing disposals may include: 1. Is radiation dose from radiological exposures either obviously negligible or obviously unacceptable based on initial modeling results? 2. If dose is not obviously negligible or unacceptable relative to applicable performance criteria, what model refinements would reduce uncertainty in the results to better inform decisions? (authors)

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

Sorption of Cationic Radionuclides on Chlorite and Sericite formed by the Hydrothermal Alteration - 18180

Chlorite and sericite formed by the hydrothermal alteration of granite distribute predominantly around the geological repository in deep underground. These minerals often cover the surface of a solid phase and fill the fracture of granite considered as the natural barrier to the migration of radionuclides leaching from radioactive waste. Therefore, to assess the practical performance of the natural barrier, it is necessary to clarify the interaction between radionuclides and these minerals. This study examined the sorption of cationic radionuclides by chlorite and sericite through batch experiments, whereby Cs{sup +}, Sr{sup 2+}, and Eu{sup 3+} were used as tracer ions. In these experiments, powder-form chlorite or sericite was mixed with the solutions of the tracer ions to a liquid/solid ratio of 10 ml/g, and the resulting mixture was allowed to incubate for 2-7 days. The concentration of tracer ions was set to 1 mM and that of NaCl to a specific value in the 0-0.6 M range. The solution pH was 3, 5, or 8. As a result, tracer ions were significantly adsorbed by chlorite and sericite, regardless of pH or NaCl concentration, except when NaCl concentration reached the high value of 0.6 M. The distribution coefficient for europium ions at pH 5 was estimated to be 2-7 mL/g in chlorite and 10{sup 3}-10{sup 4} mL/g in sericite. These values significantly exceeded that measured for feldspar (2-4 mL/g), the predominant component of granite before it undergoes any hydrothermal alteration. These results suggest that cationic radionuclides will be immobilized through sorption in chlorite and sericite filling granite fractures, even if these thermal-fluid-formed minerals cover the surface of granite. (authors)

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

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish release fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Phosphate-Mediated Remediation of Metals and Radionuclides

Worldwide industrialization activities create vast amounts of organic and inorganic waste streams that frequently result in significant soil and groundwater contamination. Metals and radionuclides are of particular concern due to their mobility and long-term persistence in aquatic and terrestrial environments. As the global population increases, the demand for safe, contaminant-free soil and groundwater will increase as will the need for effective and inexpensive remediation strategies. Remediation strategies that include physical and chemical methods (i.e., abiotic) or biological activities have been shown to impede the migration of radionuclide and metal contaminants within soil and groundwater. However, abiotic remediation methods are often too costly owing to the quantities and volumes of soils and/or groundwater requiring treatment. The in situ sequestration of metals and radionuclides mediated by biological activities associated with microbial phosphorus metabolism is a promising and less costly addition to our existing remediation methods. This review highlights the current strategies for abiotic and microbial phosphate-mediated techniques for uranium and metal remediation.

Martinez, Robert J.↗