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DOE OSTI2026

Lessons Learned from Site Selection Efforts for Deep Geological Disposal

This report discusses international experiences from site selection of deep geological repositories (DGR) for radioactive waste disposal. Based on experiences from three countries with recent or ongoing site selections, the report describes different site selection approaches and summarizes lessons learned. The German and Swiss programs are two slightly different examples of one end of the spectrum of potential siting approaches: Starting with the entire country, a technically based downselection is performed to select the repository site. The Canadian program is an example of the other end of the spectrum: Here, the program started with an extensive engagement program with Canadians to collaboratively develop a socially acceptable and technically sound national approach for the long-term management of used nuclear fuel. This was followed by the launch of a collaboratively developed consent-based siting process designed to seek an informed and willing host with suitable sites, considering both technical and community well-being site requirements.

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

Overview of Kinectrics' Experience in Characterization of Low and Intermediate Level Waste (L and ILW) - 19144

Kinectrics has over two decades of experience in physico-chemical and radiological characterization of L and ILW generated from the operation and refurbishment of nuclear reactors owned by Ontario Power Generation (OPG). This characterization is required to develop an inventory and also to support the safety case, as well as licensing, of OPG's waste management facilities including its planned Deep Geological Repository (DGR) for Low and Intermediate Level Waste (L and ILW). The typical waste matrices characterized include reactor artefacts, non-processable, compactable, incinerable, ash, assorted resin, filters, and sludge. The objective of this paper is to present an overview of the characterization work undertaken over the years and further elaborate on some recent activities pertaining to characterization of a select subset of L and ILW streams such as pressure tubes, feeder pipes, bulk non-processable and active liquid waste. (authors)

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

Recent Developments in Radioactive Waste Management in the Federal Republic of Germany - 17450

In 2013 the Federal Parliament and the Federal Council passed the Act Concerning the Search and Selection of a Site for a Repository for Heat-Generating Radioactive Waste and the Amendment of Other Acts (Site Selection Act). It was a restart for the search for a repository for especially heat-generating waste in Germany. As a first step in May 2014 the Commission 'Storage of High-Level Radioactive Waste' started their work. Their aim was to debate and clarify the principle questions concerning the management of heat-generating radioactive waste. As well as to evaluate and determine the fundamentals for the Site Selection Process e.g. safety requirements, exclusion criteria, minimum requirements and consideration criteria for the site selection. They also evaluated the organizational and procedural arrangements in the scope of waste management. As a first act to implement the suggestions of this Commission the organizational structures in waste management were greatly altered. This reorganization not only had a significant influence on the waste management of high-active, but also on the management of low- and intermediate-active waste management in the Federal Republic of Germany. In 2015 a second Commission started its work focusing on the funding of waste management in Germany. The results of this Commission were presented in April 2016. At the moment the suggestions of this Commission are being implemented, changing the financial responsibilities for waste management in the Federal Republic of Germany. The presentation will focus on changes and the developments of radioactive waste management in Germany and talk about the ongoing process in implementing the suggestions of the two Commissions. (authors)

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

Thermophysical properties of reconsolidating crushed salt.

Reconsolidated crushed salt is being considered as a backfilling material placed upon nuclear waste within a salt repository environment. In-depth knowledge of thermal and mechanical properties of the crushed salt as it reconsolidates is critical to thermal/mechanical modeling of the reconsolidation process. An experimental study was completed to quantitatively evaluate the thermal conductivity of reconsolidated crushed salt as a function of porosity and temperature. The crushed salt for this study came from the Waste Isolation Pilot Plant (WIPP). In this work the thermal conductivity of crushed salt with porosity ranging from 1% to 40% was determined from room temperature up to 300°C, using two different experimental methods. Thermal properties (including thermal conductivity, thermal diffusivity and specific heat) of single-crystal salt were determined for the same temperature range. The salt was observed to dewater during heating; weight loss from the dewatering was quantified. The thermal conductivity of reconsolidated crushed salt decreases with increasing porosity; conversely, thermal conductivity increases as the salt consolidates. The thermal conductivity of reconsolidated crushed salt for a given porosity decreases with increasing temperature. A simple mixture theory model is presented to predict and compare to the data developed in this study.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Improbability of Nuclear Criticality in Compacted Criticality Control Overpacks after Room Closure by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, nuclear criticality is improbable after salt creep causes compaction of criticality control overpacks (CCOs) disposed at the Waste Isolation Pilot Plant, an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality potential has been updated here because of the introduction of CCOs, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCOs during two representative conditions: (1) large salt block fall, and (2) gradual salt compaction (without brine seepage and subsequent gas generation to permit maximum room closure). Geomechanical models of rock fall demonstrate three tiers of CCOs are not greatly disrupted. Geomechanical models of gradual room closure from salt creep predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Criticality models of spheres and cylinders of 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular spacing demonstrate that an array of CCOs is not critical when surrounded by salt and magnesium oxide, provided the amount of hydrogenous material shipped in the CCO (usually water and plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

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

Coupled Thermo-Hydrological-Mechanical-Chemical Behavior of Anisotropic Granite for Geologic Disposal of High-Level Radioactive Waste: A Core-Scale Laboratory Investigation

The coupled thermo-hydrological-mechanical-chemical (THMC) behavior of rock within an Excavation Damaged Zone (EDZ) is critical for the safety and long-term performance of a geological repository for high-level radioactive wastes. While many laboratory experiments have been conducted to investigate EDZ rocks, the flow and deformation characteristics resulting from anisotropic rock textures and microcrack distribution under triaxial loading and elevated temperatures remain poorly understood. Particularly, cracks at various scales serve as fast paths for fluid flow and solute transport and present as focal points of mechanical weakness, which complicate the coupled THMC processes in anisotropic EDZ rocks and challenge modeling predictions. Here, in this study, a series of core-scale experiments was conducted on three granite samples under repository-relevant conditions. These rock samples were obtained from the Grimsel Underground Research Laboratory (URL), featured by anisotropic minerals (represented by bedding layers) and microcrack distributions and coarse cm-scale grain sizes. During the experiments, samples were subjected to an elevated temperature at 90 °C and different triaxial loading conditions either by radial (normal to bedding layers) or axial (parallel to bedding layers) compaction. Water was injected into the samples, and the rock permeability evolutions and effluent water chemistry were monitored closely. For intact samples, thermal expansion of minerals at 90 °C resulted in a large, 75% irreversible permeability reduction and rock strengthening under radial compaction, while thermal impact was limited to a 15% permeability reduction under axial compaction. In contrast, for a sample containing open cracks, the growth of fractures during the experiment resulted in an abrupt permeability increase and fast failure at 90 °C. The effluent water chemistry indicates much more considerable mineral dissolution from large shear sliding than that in rocks dominated by mechanical compaction. These results helped better understand the coupled THMC processes in anisotropic rocks containing cracks, evaluate the behaviors of EDZ rocks, and predict the long-term evolution of EDZ for the performance of the repositories.

Coupled THMC processes↗
DOE OSTI2018

Enhanced Evaluation Criteria to Meet the WIPP Waste Acceptance Criteria - 18655

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the disposal of Transuranic (TRU) waste and the only such disposal facility of its kind. TRU waste is radioactive mixed waste generated by the U. S. Department of Energy in support of National Defense. In 2014, the WIPP facility was shut down when two separate accidents occurred in the repository. The two accidents were a fire on a large underground haul truck that prompted evacuation and filled the mine with smoke and soot, and a radiation release that contaminated the exhaust ventilation and disposal areas of the mine. DOE performed two accident investigations in 2014 that resulted in many corrective actions. Some of the corrective actions resulted in significant changes in the WIPP Waste Acceptance Criteria (WAC) and are required to be applied to all TRU waste upon the resumption of shipments to WIPP. WIPP re-opened and began taking shipments again in 2017 and a large backlog of over 25,000 TRU waste containers that were certified prior to the new WAC have been undergoing re-evaluation since the re-opening. This paper will focus on the new WAC criteria that TRU waste must meet above-and-beyond previous WAC criteria and the implementation of new processes for approval. The primary area for these changes is in the area of waste characterization using Acceptable Knowledge (AK) and new detailed analyses required for chemical compatibility. The accident investigation revealed that the radiation release was due to a thermal runaway and bursting of the waste container due to inadvertent mixtures of nitrates and combustibles resulting in ignitable waste forms. In other words, the TRU waste container had incompatible contents that went undiscovered and created a hazard characteristic that was prohibited at WIPP. Through detailed chemical compatibility evaluations and new oversight, the DOE should be able to prevent a similar accident (see Figure 1). The new chemical compatibility evaluations and evaluations of oxidizing chemicals combined with combustibles are complex and detailed. Chemical compatibility is not a new requirement, but a much more detailed evaluation is a new requirement. And, obtaining historical information about the generation of the final waste form for legacy packaged waste can be very challenging. Specifically, determining possible amounts of chemicals in the waste can be extremely challenging. This paper does not explore in depth, the details of oxidizing chemicals. Other papers presented will discuss oxidizing chemicals. (author)

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

Thermophysical properties of the Ghareb formation relevant for nuclear waste disposal

The Ghareb formation, a shallowly buried porous chalk in Israel, is currently a candidate for nuclear waste disposal. The potential repository is somewhat unique for its host rock and emplacement in shallow (500 m) engineered large diameter boreholes. Herein, the thermal properties of the Ghareb are determined to support design and performance assessment; the relevant properties measured are thermal conductivity, specific heat capacity, thermal diffusivity, and thermal expansion coefficient along with their relationship with varying temperature. For the temperature range of 40 to 275 °C, the thermal conductivity ranges from 0.30 to 1.10 W/m·K, the thermal diffusivity ranges from 0.20 to 0.72 mm 2 /s, and volumetric heat capacity ranges from 0.86 to 2.00 MJ/m 3 ·K. Thermal strain measurements were used to estimate the linear thermal expansion coefficient to be 6·10 –4 –9·10 –2 °C -1 from 40 to 300 °C. These measured properties were used in a thermomechanical model to estimate near-field stresses an hour and 10 years after waste emplacement; the borehole was found to be stable. Thermal loading after 10 years was predicted to elevate local pore pressures by 1–1.5 MPa. The laboratory measurements coupled with analyses are the first attempts at performance assessment characterization for this first of its kind potential repository setting with this chosen host rock.

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

Verification Testing of OLI Systems Mixed Solvent Electrolyte Model for the Na-K-Mg-Ca-H-Cl-SO 4 -OH-HCO 3 -CO 3 -CO 2 -H 2 ) System to High Ionic Strength at 25°C

This technical report summarizes model verification results and summary statistics for 41 evaporite mineral solubility cases evaluated by Savannah River National Laboratory using OLI Systems’ aqueous electrolyte thermodynamic modeling software. The 41 verification cases containing a total of 60 solubility curves comprise mineral solubility data from low to high ionic strength at 25°C for the eight-component system Na-K-Mg-Ca-H-Cl-SO 4 -OH-HCO 3 -CO 3 -CO 2 -H 2 O as reported by Harvie et al. (1984). Thermodynamic calculations were executed using OLI Systems’ Stream Analyzer computation module within the OLI Studio software platform (Ver. 11.0, Rev. 11.0.1.9). The Mixed Solvent Electrolyte (MSE) thermodynamic framework was chosen for this investigation because of its superiority in modeling high ionic-strength inorganic salt solutions and actinide redox chemistry and solubility, both of which are relevant to the geological repository conditions at the Waste Isolation Pilot Plant in Carlsbad, New Mexico. Mineral solubility data in various inorganic salt solutions were digitized and extracted from figures generated by Harvie et al. (1984). For each of the 60 solubility curves, a case-specific chemistry model and input file were generated in OLI Studio using OLI Stream Analyzer and the MSE (H 3 O + ion) public databank provided by OLI Systems. Model simulation results were exported to Microsoft Excel to calculate summary statistics and to generate graphs comparing the OLI model predictions to the solubility data. Summary statistics include residuals (model – data) and concordance (accuracy × precision, where precision is indicated by the Pearson correlation coefficient and accuracy accounts for bias and scale differential). Private databanks were not developed, and activity coefficient model regressions were not performed to improve OLI model fits to the data. Of the 41 model verification plots, 83% have a mean of the percent residuals less than or equal to 25%. Similarly, 75% display a concordance greater than or equal to 0.75. Only seven of the 41 verification plots fail to show good agreement between the model and data. Of these seven, three are relevant to the WIPP repository because they involve the Mg-OH-Cl-SO 4 -CO 3 aqueous system. The remaining four address salt solubilities at the pH extremes (strong acid and strong base). It should be noted that in two of the three Mg-OH-Cl-SO 4 -CO 3 system cases, the regressed Harvie et al. (1984) solubility curve also deviated from the data. Lack of agreement between the OLI model-predicted solubility curves and the data is attributable to one or more of the following: specific solid species are not included in the OLI MSE databank; there is significant variation among the different solubility datasets chosen by Harvie et al. (1984); the OLI MSE model’s thermodynamic parameters were determined using different solubility datasets; and the activity coefficient parameters for certain relevant ion-ion and ion-molecule pairs have not been optimized via data regression. Two recommendations for future work are to (1) evaluate solubility data for the Mg-OH-Cl-SO 4 -CO 3 system at high ionic strength and, if necessary, develop a private OLI MSE database that includes missing species and, where necessary, regressed standard state properties and interaction parameters; (2) perform similar verification testing of the OLI model for actinide solubility data.

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

Exploration and Modeling of Structural changes in Waste Glass Under Corrosion

The corrosion of nuclear waste glasses is a complex process involving adsorption, diffusion, ionexchange, hydrolysis and precipitation of mineral alteration phases. These processes occur in parallel and their relative rates change significantly over the long term. Corrosion models are required to accurately reflect the long-term durability of vitrified high level waste so that geologic repositories can be designed to meet the targets of stability and security, and at the same time, make better utilization of the storage volume by increasing the density of HLW in the glass.

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

Deposition Behavior of Supersaturated Silicic Acid onto Illite as an Altered Backfilling Material - 18114

The pH gradient, which is generated by cement-based materials, causes the deposition of silicic acid in the backfilled tunnels in the repository. This deposition behavior blocks the flow-path, whence it also acts effectively for the retardation of radionuclides besides the similar effect with swelling property. Such deposition effect is expected to be maintained in the backfilling material even after illitization has progressed. The aim of this study is to evaluate the deposition rate constant, k [m/s], of supersaturated silicic acid on illite under some different conditions, considering its temperature dependence. In the experiments, deposition behavior was observed during 6 hours. After adding illite powder into the supersaturated silicic acid solution (pH>10), the pH was set to 8 in order to adjust the solution supersaturated. As a result of these experiments, the values of k were several 10-10 m/s in the temperature range of 288 K to 323 K. Several previous studies with some experiments and numerical simulations showed that the values of k for various minerals, e.g. Ca-type bentonite and amorphous silica, were in the range of 10-11 to 10-10 m/s, and that those values are large enough to clog micro flow-paths. These means that illite shows the clogging effect caused by the deposition of supersaturated silicic acid as in other silicate minerals. In addition, the deposition amount of supersaturated silicic acid increased as the temperature was raised. Therefore, the deposition reaction could progress rapidly under the condition of relatively high temperature caused by the geothermal heat and the waste heat in the repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Modeling glass degradation and release of radionuclides from vitrified waste for performance assessment simulations

The release of radionuclides initially encapsulated in a slowly degrading solid waste form and contained in an eventually corroding canister defines the source term for numerical simulations for the assessment of a geologic repository for high-level radioactive waste. While the details of waste degradation, canister corrosion, and dissolution and mobilization of the radionuclides in pore water include complex chemical reaction and transport processes that are coupled to the thermal, hydrological, microbiological, and mechanical conditions in the repository, the source-term model suitable for use in a numerical performance assessment model should be a defensible abstraction of these mechanisms. We developed a radiological source-term model and implemented it into a non-isothermal flow and transport simulator. While the proposed source-term model is applicable to various waste forms, canister systems, and disposal concepts, we specifically considered radionuclide releases from vitrified high-level waste placed in a cylindrical canister disposed in a deep vertical borehole repository. In this model, waste degradation is a function of temperature, and it can be adjusted to evaluate the influence of and propagate uncertainties in pH, passivation reactions, and chemical conditions as well as geometrical factors. The time-dependent, congruent release of safety-relevant radionuclides present in the decaying inventory is then calculated. Finally, the radionuclides are mobilized by diffusive and advective transport according to the thermo-hydraulic conditions prevailing in the near field of the repository, from where they migrate through the geosphere to the accessible environment. We examine the influence of the source-term model’s parameters on performance assessment calculations through sensitivity and uncertainty propagation analyses, identifying influential factors and confirming the upper bound of their impact. These considerations align with the overarching goal of repository design, which is to demonstrate that engineered and natural barriers can collectively delay radionuclide migration for timescales far exceeding human planning, thereby providing multiple, redundant barriers against environmental contamination.

iTOUGH2↗
DOE OSTI2016

Implications of Grain-Scale Mineralogical Heterogeneity for Radionuclide Transport in Fractured Media

The geological disposal of nuclear waste is based on the multi-barrier concept, comprising various engineered and natural barriers, to confine the radioactive waste and isolate it from the biosphere. Some of the planned repositories for high-level nuclear waste will be hosted in fractured crystalline rock formations. The potential of these formations to act as natural transport barriers is related to two coupled processes: diffusion into the rock matrix and sorption onto the mineral surfaces available in the rock matrix. Different in situ and laboratory experiments have pointed out the ubiquitous heterogeneous nature of the rock matrix: mineral surfaces and pore space are distributed in complex microstructures and their distribution is far from being homogeneous (as typically assumed by Darcy-scale coarse reactive transport models). We use a synthetically generated fracture–matrix system to assess the implications of grain-scale physical and mineralogical heterogeneity on cesium transport and retention. The resulting grain-scale reactive transport model is solved using high-performance computing technologies, and the results are compared with those derived from two alternative models, denoted as upscaled models, where mineral abundance is averaged over the matrix volume. In the grain-scale model, the penetration of cesium into the matrix is faster and the penetration front is uneven and finger-shaped. The analysis of the cesium breakthrough curves computed at two different points in the fracture shows that the upscaled models provide later first-arrival time estimates compared to the grain-scale model. The breakthrough curves computed with the three models converge at late times. These results suggest that spatially averaged upscaled parameters of sorption site distribution can be used to predict the late-time behavior of breakthrough curves but could be inadequate to simulate the early behavior.

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

Solidification of Low and Intermediate Level Waste - A Comparison between Different Methods - 19584

The treatment of radioactive materials results in products that have to be suitable for a period of interim storage as well as for disposal in a repository. Considering available options for treatment of radioactive waste, future disposal also has to be taken into account. Up to now, only cementation and encapsulation in geopolymer (only to limited extent) are relevant treatment methods for the solidification of German waste for disposal in the repository Grube Konrad. However, a synopsis of other solidification technologies is considered to be helpful in order to be prepared for future developments in the fields of radioactive waste treatment. Treatment methods addressed in this paper are: - Cementation; - Encapsulation into geopolymer matrix; - Vitrification; - SYNROC; - Geomelting. The criteria for evaluation of the treatment methods are as follows: - Requirements concerning waste properties; - Safety of treatment facilities; - Fulfillment of waste acceptance criteria for repository Grube Konrad; - Fulfillment of requirements for near surface disposal; - Suitability for long-term interim storage. (authors)

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

Reinvestigation into Closure Predictions of Room D at the Waste Isolation Pilot Plant

Room D was an in-situ, isothermal, underground experiment conducted at theWaste Isolation Pilot Plant between 1984 and 1991. The room was carefully instrumented to measure the horizontal and vertical closure immediately upon excavation and for several years thereafter. Early finite element simulations of salt creep around Room D under predicted the vertical closure by 4.5×, causing investigators to explore a series of changes to the way Room D was modeled. Discrepancies between simulations and measurements were resolved through a series of adjustments to model parameters, which were openly acknowledged in published reports. Interest in Room D has been rekindled recently by the U.S./German Joint Project III and Project WEIMOS, which seek to improve the predictions of rock salt constitutive models. Joint Project participants calibrate their models solely against laboratory tests, and benchmark the models against underground experiments, such as room D. This report describes updating legacy Room D simulations to today’s computational standards by rectifying several numerical issues. Subsequently, the constitutive model used in previous modeling is recalibrated two different ways against a suite of new laboratory creep experiments on salt extracted from the repository horizon of the Waste Isolation Pilot Plant. Simulations with the new, laboratory-based, calibrations under predict Room D vertical closure by 3.1×. A list of potential improvements is discussed.

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

Reinvestigation into Closure Predictions of Room D at the Waste Isolation Pilot Plant

Room D was an in-situ, isothermal, underground experiment conducted at the Waste Isolation Pilot Plant between 1984 and 1991. The room was carefully instrumented to measure the horizontal and vertical closure immediately upon excavation and for several years thereafter. Early finite element simulations of salt creep around Room D under-predicted the vertical closure by 4.5×, causing investigators to explore a series of changes to the way Room D was modeled. Discrepancies between simulations and measurements were resolved through a series of adjustments to model parameters, which were openly acknowledged in published reports. Interest in Room D has been rekindled recently by the U.S./German Joint Project III and Project WEIMOS, which seek to improve the predictions of rock salt constitutive models. Joint Project participants calibrate their models solely against laboratory tests, and benchmark the models against underground experiments, such as room D. This report describes updating legacy Room D simulations to today’s computational standards by rectifying several numerical issues. Subsequently, the constitutive model used in previous modeling is recalibrated two different ways against a suite of new laboratory creep experiments on salt extracted from the repository horizon of the Waste Isolation Pilot Plant. Simulations with the new, laboratory-based, calibrations under-predict Room D vertical closure by 3.1×. A list of potential improvements is discussed.

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

Multistage bioassociation of uranium onto an extremely halophilic archaeon revealed by a unique combination of spectroscopic and microscopic techniques

The interactions of two extremely halophilic archaea with uranium were investigated in this paper at high ionic strength as a function of time, pH and uranium concentration. Halobacterium noricense DSM-15987 and Halobacterium sp. putatively noricense, isolated from the Waste Isolation Pilot Plant repository, were used for these investigations. The kinetics of U(VI) bioassociation with both strains showed an atypical multistage behavior, meaning that after an initial phase of U(VI) sorption, an unexpected interim period of U(VI) release was observed, followed by a slow reassociation of uranium with the cells. By applying in situ attenuated total reflection Fourier-transform infrared spectroscopy, the involvement of phosphoryl and carboxylate groups in U(VI) complexation during the first biosorption phase was shown. Differences in cell morphology and uranium localization become visible at different stages of the bioassociation process, as shown with scanning electron microscopy in combination with energy dispersive X-ray spectroscopy. Finally, our results demonstrate for the first time that association of uranium with the extremely halophilic archaeon is a multistage process, beginning with sorption and followed by another process, probably biomineralization.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Krypton-81 in groundwater of the Culebra Dolomite near the Waste Isolation Pilot Plant, New Mexico

TheWaste Isolation Pilot Plant (WIPP) in New Mexico is the first geologic repository for disposal of transuranic nuclear waste fromdefense-related programs of the US Department of Energy. It is constructed within halite beds of the Permian-age Salado Formation. The Culebra Dolomite, confined within Rustler Formation evaporites overlying the Salado Formation, is a potential pathway for radionuclide transport from the repository to the accessible environment in the human-disturbed repository scenario. Although extensive subsurface characterization and numerical flow modeling of groundwater has been done in the vicinity of theWIPP, few studies have used natural isotopic tracers to validate the flow models and to better understand solute transport at this site. The advent of Atom-Trap Trace Analysis (ATTA) has enabled routine measurement of cosmogenic 81Kr (half-life 229,000 yr), a near-ideal tracer for long-term groundwater transport. We measured 81Kr in saline groundwater sampled from two Culebra Dolomite monitoring wells near the WIPP site, and compared 81Kr model ages with reverse particle-tracking results of well-calibrated flow models. The 81Kr model ages are ~130,000 and ~330,000 yr for high-transmissivity and low-transmissivity portions of the formation, respectively. Compared with flow model results which indicate a relatively young mean hydraulic age (~32,000 yr), the 81Kr model ages imply substantial physical attenuation of conservative solutes in the Culebra Dolomite and provide limits on the effective diffusivity of contaminants into the confining aquitards.

Sturchio, Neil C.↗