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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Neptunium(IV) Diffusion through Bentonite Clay

DOE Fellow, Silvina A. Di Pietro, completed a 10-week internship at Lawrence Livermore National Laboratory (LLNL) in Livermore, California. The objective of the study was to determine the diffusion profile of neptunium (Np) in bentonite clay (potential geo-engineered barrier for nuclear waste repositories). Under the mentorship of Drs. Claudia Joseph (Karlsruhe Institute of Technology, Germany) and Mavrik Zavarin, she prepared a series of core abrasion peeling samples. This procedure was a continuation of a Np diffusion experiment conducted by Dr. Joseph during her post-doc residency at LLNL in the year 2016. The second part of the experiment (from now on will be referred as abrasive peeling) focuses on understanding (1) the Np movement through the clay core and (2) the influence of temperature (25°C, 65°C, and 85°C) for the Np transport across the clay core under anaerobic conditions. End of 2016, the diffusion core samples were embedded in epoxy resin to stabilize them for long-term. The abraded core powder samples vary in material (epoxy vs. clay), Np activity and concentration, temperature exposed during the diffusion experiment, and mass, or amount of clay peeled off. Throughout the course of the 10-week internship program at LLNL, the experiments implemented and discussed in this report aimed to answer the questions: (1) How is the Np(IV) diffusion profile through bentonite influenced by the temperature? and (2) Will Np(IV) migrate through the geoengineered barrier of a nuclear waste repository? In addition, Ph.D. candidate Di Pietro received invaluable guidance in experimental design and development that will benefit her ongoing research endeavors.

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

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

The Use of Positrons to Survey Alteration Layers on Synthetic Nuclear Waste Glasses

In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe0. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

Doppler Broadening Positron Annihilation Spectrosc↗
DOE OSTI2020

Near-Field Corrosion Interactions between Glass and Corrosion Resistant Alloys

This study explores the corrosion interactions between model nuclear waste glass materials and corrosion resistant alloys, under conditions that simulate the near field of a nuclear waste repository. The interactions between the corrosion of stainless steel (SS) 316, alloy G30, or alloy 625, and international simple glass or soda-lime silica glass are systematically studied. The dissimilar materials were exposed in close proximity to each other in different electrolytes at 90 oC. After exposure, the glass surface exposed near metals showed different regimes of corrosion, with distinct surface morphologies and chemistries that were likely affected by the local environment created by the localized corrosion of metals. Surface and solution analyses showed that the corrosion rate of glass was enhanced by the presence of metals. Infrared spectroscopy data suggested the local build-up of stresses in the contact area of glass, which may lead to the mechanical instability of the glass alteration layer. On the other hand, the effect of glass on metal corrosion is strongly dependent on the leaching solution. In electrolytes containing abundant aggressive anions such as Cl-, glass seems to suppress the localized corrosion of SS by the precipitation of a Si-rich surface film that protects the SS substrate from solutions. However, in less aggressive electrolytes, the corrosion rate of SS was increased by the presence of glass corrosion products. Overall, our study showed that the hidden and localized damage on glass in contact with metals may enhance the release of incorporated radionuclides compared to typical uniform glass corrosion. Therefore, more attention should be paid to the corrosion interactions between metals and glass under repository conditions.

Guo, Xiaolei↗
DOE OSTI2022

Phenomenological Model of Nonlinear Dynamics and Deterministic Chaotic Gas Migration in Bentonite: Experimental Evidence and Diagnostic Parameters

Understanding gas migration in compacted clay materials, e.g., bentonite and claystone, is important for the design and performance assessment of an engineered barrier system of a radioactive waste repository system, as well as many practical applications. Existing field and laboratory data on gas migration processes in low-permeability clay materials demonstrate the complexity of flow and transport processes, including various types of instabilities, caused by nonlinear dynamics of coupled processes of liquid–gas exchange, dilation, fracturing, fracture healing, etc., which cannot be described by classical models of fluid dynamics in porous media. We here show that the complexity of gas migration processes can be explained using a phenomenological concept of nonlinear dynamics and deterministic chaos theory. To do so, we analyzed gas pressure and gas influx (i.e., input) and outflux (i.e., output), recorded during the gas injection experiment in the compact Mx80-D bentonite sample, and calculated a set of the diagnostic parameters of nonlinear dynamics and chaos, such a global embedding dimension, a correlation dimension, an information dimension, and a spectrum of Lyapunov exponents, as well as plotted 2D and 3D pseudo-phase-space strange attractors, based on the univariate influx and outflux time series data. These results indicate the presence of phenomena of low-dimensional deterministic chaotic behavior of gas migration in bentonite. In particular, during the onset of gas influx in the bentonite core, before the breakthrough, the development of gas flow pathways is characterized by the process of chaotic gas diffusion. After the breakthrough, with inlet-to-outlet movement of gas, the prevailing process is chaotic advection. During the final phase of the experiment, with no influx to the sample, the relaxation pattern of gas outflux is resumed back to a process of chaotic diffusion. The types of data analysis and a proposed phenomenological model can be used to establish the basic principles of experimental data-gathering, modeling predictions, and a research design.

36 MATERIALS SCIENCE↗
DOE OSTI2024

M3SF-24LL010302062-NEA-TDB Management and International Collaborations in Sorption and Thermodynamic Modeling

This progress report (Level 3 Milestone Number M3SF-24LL010302062) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline International Collaborations Work Package Number SF-24LL01030206. The activity is focused on our long-term commitment to engaging our partners in international nuclear waste repository research. This includes participation in the Nuclear Energy Agency Thermochemical Database (NEA-TDB) Project and development of methodologies for integrating US and international thermodynamic databases for use in SFWST Generic Disposal System Assessment (GDSA) efforts. A continuing focus for FY24 efforts is to support the US participation in the NEA-TDB effort. The focus of FY24 activities was the development of an agreement for a Phase 7 activity that will start in Q1 of 2025. Mavrik Zavarin is now the US representative on both the Management Board and the Executive group to the NEA-TDB. He is also the POC for the Cements State of the Art Report that is undergoing peer review in FY24. In FY24, we used our position on the NEA-TDB MB and EG to facilitate the integration of NEA-TDB thermochemical data with LLNL’s SUPCRTNE thermodynamic database that supports the SFWST GDSA activities. This effort is coordinated with the Argillite work package SUPCRTNE database development efforts (Wolery, 2024). The goal is to provide a downloadable database that will be hosted on LLNL’s thermodynamics website which incorporates NEA-TDB data into the LLNL database where appropriate. We also began engagement with the EURAD2 program that was initiated in FY24 by our European collaborators at the Karlsruhe Institute of Technology (KIT), Germany. The primary focus of the engagement is with WP20: DITUSC Thermodynamic database evaluation program. A kickoff meeting for this activity is planned for early FY25. Finally, we have been selected to co-host (with Clemson University) the International Conference on Chemistry and Migration Behaviour of Actinides and Fission Products in the Geosphere in 2025 (Migration2025). The meeting will be held September 21-26, 2025, in New Orleans, Louisiana, and will focus on international efforts to understand the risks of radionuclide releases into the environment. This central focus of this conference is on international efforts to develop safe disposal options for nuclear wastes. As such, we are developing a theme focused on US underground nuclear waste repository science.

58 GEOSCIENCES↗
DOE OSTI2015

The role of independent monitoring in maintaining community support through a radiological incident - a WIPP case study - 15107

The Waste Isolation Pilot Plant (WIPP) is a U.S. Department of Energy (DOE) waste repository for the permanent disposal of defense transuranic (TRU) wastes. The repository is mined in the Salado Formation, a bedded salt formation of the Permian era, approximately 655 m (2150 ft.) below land surface. Located near Carlsbad, New Mexico, the WIPP facility is the world's first deep geological disposal facility licensed to accept TRU waste. Environmental monitoring is a key component in the development and operation of any nuclear facility. Well after the WIPP facility had been sited and constructed, but before repository operations began, DOE and local community leaders recognized the value of having an independent environmental monitoring program in addition to the self-monitoring program required by regulation. An independent monitoring organization, the Carlsbad Environmental Monitoring and Research Center (CEMRC) was created by New Mexico State University at its Carlsbad campus. After almost fifteen years of successful waste disposal operations, the first airborne radiation was accidentally released from the WIPP and detected beyond the site boundary on February 14, 2014. The accident released moderate levels of radioactivity into the underground air. A small but measurable amount of radioactivity also escaped to the surface and was detected beyond the site's inner boundary. The results were reported to the public as they were obtained and verified. Releases were low and localized, and no radiation-related health effects among local workers or the public would be expected. CEMRC's independence and its extensive monitoring program played a crucial role in timely informing the local populations on the levels and types of radiation released from the WIPP. CEMRC also helped eliminate unnecessary concern surrounding fear of radiation by engaging local populations through news releases and presentations at 'town hall' meetings sponsored by the Mayor's Office and the Department of Energy. (authors)

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

International Approaches for Nuclear Waste Disposal in Geological Formations: Geological Challenges in Radioactive Waste Isolation—Fifth Worldwide Review

The overall objective of the Fifth Worldwide Review (WWR-5) is to document the current state-of-the-art of major developments in a number of nations throughout the World pursuing geological disposal programs, and to summarize challenging problems and experience that have been obtained in siting, preparing and reviewing cases for the operational and long-term safety of proposed and operating nuclear waste repositories. The scope of the Review is to address current specific technical issues and challenges in safety case development along with the interplay of technical feasibility, siting, engineering design issues, and operational and post-closure safety. In particular, the chapters included in the report present the following types of information: the current status of the deep geological repository programs for high level nuclear waste and low- and intermediate level nuclear waste in each country, concepts of siting and radioactive waste and spent nuclear fuel management in different countries (with the emphasis of nuclear waste disposal under different climatic conditions and different geological formations), progress in repository site selection and site characterization, technology development, buffer/backfill materials studies and testing, support activities, programs, and projects, international cooperation, and future plans, as well as regulatory issues and transboundary problems.

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

Radionuclide transport in fractured chalk under abrupt changes in salinity

Internationally, it has been agreed that geologic repositories for spent fuel and radioactive waste are considered the internationally agreed upon solution for intermediate and long-term disposal. In countries where traditional nuclear waste repository host rocks (e.g., clay, salt, granite) are not available, other low permeability lithologies must be studied. Here, chalk is considered to determine its viability for disposal. Despite chalk's low bulk permeability, it may contain fracture networks that can facilitate radionuclide transport. In arid areas, groundwater salinity may change seasonally due to the mixing between brackish groundwater and fresh meteoric water. Such salinity changes may impact the radionuclides' mobility. In this study, radioactive U(VI) and radionuclide simulant tracers (Sr, Ce and Re) were injected into a naturally fractured chalk core. The mobility of tracers was investigated under abrupt salinity variations. Two solutions were used: a low ionic strength (IS) artificial rainwater (ARW; IS ~0.002) and a high IS artificial groundwater (AGW; IS ~0.2). During the experiments, the tracers were added to ARW, then the carrier was changed to AGW, and vice versa. Ce was mobile only in colloidal form, while Re was transported as a conservative tracer. Both Re and Ce demonstrated no change in mobility due to salinity changes. In contrast, U and Sr showed increased mobility when AGW was introduced and decreased mobility when ARW was introduced into the core. These experimental results, supported by reactive transport modeling, suggest that saline groundwater solutions promote U and Sr release via ion-exchange and enhance their migration in fractured chalk. Here the study emphasizes the impact of salinity variations near spent fuel repositories and their possible impact on radionuclide mobility.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

Feasibility of Using Steel Drums for the Confinement of Burned Low-Level Radioactive Waste Ashes

The selection of suitable materials to construct a facility for the final disposal of radioactive wastes (repository) is a challenge in materials science, since these facilities must be designed to provide confinement to the radionuclides for long periods of time. In the case of the low-level radioactive waste (LLRW) repository, the confinement of the species must be guaranteed for more than 300 years. In this context, the evaluation of the proposed materials must be performed under accelerated laboratory tests. One type of LLRW consists of ashes that result from the incineration of solid combustible radioactive waste. In Argentina, one of the proposed projects is to place these ashes inside steel drums and immobilized them by a cementation process. The issue to be resolved is whether the steel drums will undergo internal corrosion depending on the presence of certain deleterious species. The aim of the present work is to evaluate the corrosion susceptibility of steel drums containing simulated cemented incineration ashes. To this purpose, a special type of specimen was designed and the corrosion process was followed up for almost 4 years. It was found that, although the deleterious effect of chloride is undeniable, corrosion rates are not as high as to compromise the integrity of the steel drums for a design period of 300 years.

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

Simulating the Effect of Fracture Connectivity on Repository Performance with GDSA Framework - 18589

This work examines the possible migration of radionuclides from a deep geologic nuclear waste repository sited in fractured crystalline host rock. The key safety concern is the potential effect on waste isolation of the inter-connectivity of the fracture network, which is primarily established by the temporal evolution of the temperature and stress fields at the time of original rock intrusion and cooling. Two end members are considered in these simulations, one with a high degree of connectivity to the biosphere, such that advective transport through the fracture and fault network controls radionuclide migration, and the other with a low degree of connectivity, such that slow diffusive transport through the crystalline rock matrix is the controlling process for radionuclide migration to the biosphere. Both end members have several areally extensive, high-transmissivity deformation zones, whose distance to the repository is an important factor for waste isolation capability. Uncertainties in fracture properties (transmissivity, orientation, radius) and fracture distribution also give rise to uncertainty in waste isolation capability (i.e., the overall connectivity of the repository with the surface biosphere). These types of natural system uncertainties are generally present in any geologic site-characterization program and thus represent an important factor in assessing repository performance in hard rock environments. The waste considered in these simulations are spent fuel rod assemblies from the U.S. commercial nuclear reactor fleet. Thermal output of the waste must be considered in the simulations and can influence the rate and timing of waste package failure, waste form degradation, and fluid flux due to thermal expansion around the repository horizon. To achieve a high-fidelity representation of radionuclide transport in fractures and rock matrix, combined with thermal energy transport and fluid flow in fractures and matrix, the mathematical model is solved numerically in a parallel high-performance computing (HPC) environment on a finite volume unstructured grid consisting of approximately 5 million cells, using Geologic Disposal Safety Assessment (GDSA) Framework (https://pa.sandia.gov), an open-source performance assessment tool for deep underground disposal of nuclear waste. GDSA Framework uses PFLOTRAN to solve the balance equations on a three-dimensional grid with heterogeneous properties, using multiple processors in a parallel configuration based on domain decomposition. The fracture networks in these simulations are originally generated as discrete fracture networks (DFNs), which are sets of two-dimensional planes distributed in a three-dimensional domain. The method used in GDSA Framework maps the stochastically generated DFN to an equivalent continuous porous medium (ECPM) domain that allows for the simulation of coupled heat flow, fluid flow, and radionuclide transport, including heat conduction through the matrix of the fractured rock, which is a process not easily modeled in a DFN representation. Computational efficiency is also greatly enhanced using the ECPM method, allowing for a realistic representation and analysis of uncertainties in a multi-realization performance assessment of a deep geologic repository. The effect of fracture connectivity on the waste isolation safety function, as brought to light by these GDSA Framework simulations, points to the importance of including a realistic representation of uncertainties in fracture properties and distribution (effectively, uncertainty in spatial heterogeneity) in repository safety assessment simulations. (authors)

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

Vitrification Process for Idaho HLW Calcine - 18433

The majority of the liquid wastes produced from reprocessing of several types of spent nuclear fuel at the Department of energy's (DOE) Idaho site have been volume reduced by calcination. The resulting ∼155,600 ft{sup 3} of calcined waste is stored in 43 shielded stainless steel bins, which have a 500-year design life. The calcined material is both a high-level waste and a RCRA hazardous waste, which complicates treatment and disposal options. In addition to a potential direct disposal option, treatment options that have been considered include steam reforming, hot isostatic pressing (HIP), and vitrification. This paper describes testing and evaluation of a vitrification option for the calcine. Vitrification to produce a borosilicate glass has the advantage that it is the only treatment that provides a waste form that meets the current HLW repository waste acceptance criteria. It is therefore the lowest risk option in terms of regulatory and legislative concerns. The present work addressed several key questions: (i) Calcine waste loadings in glass, which determine the number of HLW canisters produced and the required glass production rate (throughput) for the system design; (ii) Melt rate per unit glass pool surface area of the melter, which determines the sizing of the vitrification system; and (iii) The ability to meet waste form acceptance requirements (PCT and TCLP) while achieving reasonable waste loadings. These are issues of fundamental importance to the sizing estimates for the unit operations and overall flow-sheet and they, therefore, directly impact the cost and schedule estimates. Fully compliant, high-waste-loading glasses were developed for two of the four Idaho calcine wastes. The results demonstrate that waste loadings of 35 wt% waste oxides are achievable for the two alumina calcines and about 30 wt% is achievable for the two zirconia calcines. Testing of these formulations in a joule-heated ceramic melter showed glass production rates of over 5000 kg/(m2/d) for the Alumina calcine and over 8500 kg/(m{sup 2}/d) for the Zirconium-Na Blend calcine. These extremely high production rates translate directly into smaller system size and shorter mission duration and, therefore, reduced costs. Based on these results, a preliminary design for the full-scale vitrification system was developed. The system consists of waste receipt, vitrification, off-gas treatment, and canister sealing and decontamination equipment. The calcine is blended with the appropriate glass forming chemicals (or glass frit) and fed to the melter as a solid stream. The melter has a glass pool surface area of about 2.5 m{sup 2} and is equipped with a bubbling system to increase processing rates. Glass production rates of about 6 metric tons per day are projected. The vitrification system would be installed into the existing Idaho Waste Treatment Unit Facility (IWTU) so that no new facilities are required to house the treatment system, thus greatly reducing overall cost, duration, and risk. The system would be designed to treat all of the calcine waste within a period of 12 years at an operating efficiency of 80%, producing a total of about 8500 canisters each weighing two metric tons. (authors)

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

Long-term diffusion of U(VI) in bentonite: Dependence on density

As a contribution to the safety assessment of nuclear waste repositories, U(VI) diffusion through the potential buffer material MX-80 bentonite was investigated at three clay dry densities over six years. Synthetic MX-80 model pore water was used as background electrolyte. Speciation calculations showed that Ca 2 UO 2 (CO 3 ) 3 (aq) was the main U(VI) species. The in- and out-diffusion of U(VI) was investigated separately. U(VI) diffused about 3 mm, 1.5 mm, and 1 mm into the clay plug at ρ = 1.3, 1.6, and 1.9 g/cm 3 , respectively. No through-diffusion of the U(VI) tracer was observed. However, leaching of natural uranium contained in the clay occurred and uranium was detected in all receiving reservoirs. As expected, the effective and apparent diffusion coefficients, D e and D a , decreased with increasing dry density. The D a values for the out-diffusion of natural U(VI) were in good agreement with previously determined values. Surprisingly, D a values for the in-diffusion of U(VI) were about two orders of magnitude lower than values obtained in short-term in-diffusion experiments reported in the literature. Some potential reasons for this behavior that were evaluated are changes of the U(VI) speciation within the clay (precipitation, reduction) or changes of the clay porosity and pore connectivity with time. By applying Archie's law and the extended Archie's law, it was estimated that a significantly smaller effective porosity must be present for the long-term in-diffusion of U(VI). Finally, the results suggest that long-term studies of key transport phenomena may reveal additional processes that can directly impact long-term repository safety assessments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

Hydration structure and energetics of iron-exchanged montmorillonite

The interaction between iron released from corroded steel canisters and bentonite is a key process influencing the long-term performance of nuclear waste repositories. In particular, the migration of Fe²⁺ into montmorillonite (Mnt) interlayers may alter its hydration, swelling, and ion-transport properties. In the present study, molecular dynamics simulations were performed to investigate the hydration behavior, structural response, and transport properties of Fe-exchanged montmorillonite (Fe-Mnt) under varying hydration states. The simulations focus on short- to intermediate-time-scale Fe 2+ and Fe 3+ interlayer exchange and hydration effects, and do not consider long-term structural substitution, Fe-bearing clay phase stabilization, or secondary iron mineral precipitation. Systems containing Na + -, Fe 2+ -, and Fe 3+ -Mnt were examined using both periodic and edge-exposed configurations to evaluate interlayer structure, ion exchange, and free energy of Fe intercalation. The results show that Fe ions influence the interlayer spacing primarily at low water contents (<1 bilayer), where Fe-Mnt exhibits a d-spacing 1–2 Å larger than Na-Mnt due to stronger hydration. The calculated hydration energies follow the order Fe 2+ $<$Fe 3+ $<$Na + . Both water and ion diffusion coefficients decrease upon Fe ion intercalation, with Fe 2+ ions diffusing an order of magnitude more slowly than those of Na + . Free energy profiles further confirm that Fe 2+ and Fe 3+ ions are thermodynamically favored in the interlayer, with Fe 3+ being the most stable. In conclusion, these findings provide molecular-scale insights into the mechanisms of Fe–Na exchange and their implications for bentonite alteration in repository environments.

clay↗
DOE OSTI2022

A reinvestigation into Munson’s model for room closure in bedded rock salt

Accurate predictions of room closure are important for hazardous waste repositories in rock salt formations, such as the Waste Isolation Pilot Plant (WIPP). When Munson and co-workers simulated several room closure experiments conducted at the WIPP during the 1980’s and 1990’s, their simulated closure curves closely agreed with the closure measurements. A careful review of their work, however, raised concerns and prompted the reinvestigation in this paper. Here, to begin the reinvestigation, Munson’s legacy Room D closure simulation was reasonably recreated in a current-day finite element code. Next, special care was taken to obtain numerically converged results, re-introduce the anhydrite strata intermittently ignored by Munson, and calibrate the Munson–Dawson (M–D) constitutive model for salt as much as possible from laboratory test measurements. When this new model was used to simulate Room D’s closure, it under-predicted the horizontal and vertical closure rates by 2.34x and 3.10x, respectively, at 5.7 years after room excavation. As a result, the M–D model was extended to capture the newly established creep behavior at low equivalent stresses (<8 MPa) and replace the Tresca with the Hosford equivalent stress. Simulations using the new M–D model over-predicted the horizontal closure rate by 1.15x and under-predicted the vertical closure rate by 1.08x at 5.7 years, averaged over three room closure experiments. Although further improvements could be made, the new model has a stronger scientific foundation than Munson’s legacy model and appears ready for careful engineering use.

58 GEOSCIENCES↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗