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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Effect of Anoxic Iron Corrosion on WIPP Brine Geochemistry FY23 Final Report (U)

A 280-day study was completed to evaluate the effect of zero-valent iron (Fe 0 ) on the Waste Isolation Pilot Plant (WIPP) brine geochemistry under anticipated reducing conditions. Hydrogen (H 2 ) gas is expected to be present in the repository after closure due to the anoxic corrosion of a vast quantity of iron contained in the waste forms disposed at WIPP; therefore, a background argon atmosphere containing H 2 was chosen for this study. WIPP groundwater brine pH and E h will impact the mobility and fate of plutonium within the repository. Modeling and laboratory results for Castile WIPP brine indicate that equilibrium E h values relative to the standard hydrogen electrode (SHE) are 40 mV more reducing (i.e., more negative) than those for Salado WIPP brine (-480 mV vs. -440 mV, respectively) because of the higher pH of the Castile brine (pH 9.3 for Castile vs. pH 8.8 for Salado). The E h and pH data were corrected for the effects of high ionic strength. The experimental results for both brines are consistent with thermodynamic predictions using OLI Systems' Mixed Solvent Electrolyte chemical equilibrium model. The measured and corrected pH and E h data from this study are provided in Table ES-1 and Table ES-2, respectively.

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

Radiological characterization of high Cs-137 waste items at LANL combining Cs-137 NDA results and acceptable process knowledge - 15592

Radiological characterization is an important component to disposition and disposal of radioactive waste at the Los Alamos National Laboratory. Gamma and neutron based non-destructive assay systems are available at LANL's TA-54 Area G for radiological characterization and are generally very effective in quantifying the transuranic radionuclide concentration required for a TRU or low level waste disposition determination. This is routinely performed by scaling some nuclides to direct assay results using acceptable process knowledge scaling from a directly measured key TRU nuclide, e.g., for gamma systems using a standard isotopic distribution to determine the amount of Pu-240 associated with a measured value of Pu-239. However, for two groups of LANL waste items that have high levels of Cs-137, these non-destructive assay systems were unable to make a standard TRU/LLW determination. Therefore, it was necessary to apply an extension of the typical scaling technique that keyed in on the direct Cs-137 gamma assay result to effectively characterize these waste items for TRU processing or LLW disposition. Two basic types of waste packages contained the high Cs-137 waste requiring special scaling: drums and boxes. The drums included 55-gallon, 85-gallon and 110-gallon volume sizes and the boxes included a corrugated metal box, a Standard Waste Box, and several overpack boxes containing oversized waste. All of these items contained radioactive waste generated from operations at the Chemistry and Metallurgical Research facility at LANL that included hot cell, metallographic sectioning and maintenance activities. All of the drums as well as the CMB, SWB and one oversized item in a plywood box were located above-ground at LANL's TA-54 Area G, while five of the oversized items were hot cell liners in metal boxes stored below-ground in shafts. All of the items were measured in the 2013-2014 time period at TA-54 Area G with LANL NDA systems. Most of the items were counted with LLW certified far-field gamma spectroscopy systems employing portable far-field HPGe detectors and carts and some of the items were counted on WIPP certified gamma only or gamma with neutron NDA systems. In all cases Cs-137 was quantified directly with a gamma measurement but the key TRU nuclides were not detectable. Thus, for each waste item AK/PK was studied and scaling factors to Cs-137 were established and applied to provide a comprehensive radiological characterization. A total of 49 waste items were successfully characterized as LLW based on this technique. The above ground waste items were dispositioned and disposed of at an offsite LLW repository and the below ground boxes were left in place as LLW in the TA-54 Area G shafts. This NDA with scaling application has proved effective and defensible in these high Cs-137 cases at LANL and will continue to be used as applicable and appropriate for future waste characterization at LANL. (authors)

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

Using Atom Dynamics to Map the Defect Structure Around an Impurity in Nano-Hematite

The geochemical cycling of actinides such as U is strongly controlled by Fe (oxyhydr)oxides. Indeed, the strong affinity of these minerals for U has factored into the design of geologic repositories for the long-term storage of nuclear waste. Many decades of work have focused on detailing the local coordination environment U associated with hematite (a-Fe2O3), a common and stable Fe(III) oxide, in order to better predict the fate and transport of U in the sub-surface. Here, for the first time, molecular dynamic simulations and atomically resolved electron microscopy of uranium atom mobility under the beam were used to map the topology of defect clusters surrounding structurally incorporated U in hematite. The ability to observe vacancies by means of metal atom hopping can be used to directly probe defects in other materials, with promising applications in geochemistry and materials science.

Ilton, Eugene S.↗
DOE OSTI2024

Analytical capabilities for iodine detection: Review of possibilities for different applications

This Review summarizes a range of analytical techniques that can be used to detect, quantify, and/or distinguish between isotopes of iodine (e.g., long-lived 129 I, short-lived 131 I, stable 127 I). One reason this is of interest is that understanding potential radioiodine release from nuclear processes is crucial to prevent environmental contamination and to protect human health as it can incorporate into the thyroid leading to cancer. It is also of interest for evaluating iodine retention performances of next-generation iodine off-gas capture materials and long-term waste forms for immobilizing radioiodine for disposal in geologic repositories. Depending upon the form of iodine (e.g., molecules, elemental, and ionic) and the matter state (i.e., solid, liquid, and gaseous), the available options can vary. In addition, several other key parameters vary between the methods discussed herein, including the destructive vs nondestructive nature of the measurement process (including in situ vs ex situ measurement options), the analytical data collection times, and the amount of sample required for analysis.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Thermodynamic parameters for the complexation of technetium(IV) with EDTA

Technetium-99 is a high yield (~6% fission yield) fission product and long-lived (2.13×10 5 year half-life) component of nuclear waste that will be disposed of in a geological repository. Some 99 Tc has been released into the environment due to nuclear fuel and weapon production activities at sites such as Hanford, WA. Strongly complexing ligands such as ethylenediamine- N , N , N ', N '-tetraacetic acid (EDTA) are known to increase Tc(IV) solubility and mobility in environmental systems and an accurate quantification of the complexation of Tc(IV) with EDTA is important for predicting its behavior in a geological repository. A liquid–liquid extraction system utilizing 0.2 M TOPO in dodecane was used to measure the stability constants of Tc(IV)-EDTA in 0.50 m NaNO 3 at variable temperatures (14.0±0.1, 25.0±0.1, and 32.0±0.1°C). The acid dependence of the apparent stability constants in the p C H range of 2.00–2.70 indicated the formation of TcO(EDTA) 2- (log β 101 =17.9±0.3, 25.0±0.1°C) and a protonated complex TcO(H)(EDTA) - (log β 111 =20.5±0.1, 25.0±0.1°C). The associated thermodynamic parameters Δ r G 101 =-101.7±0.4 kJ·mol -1 , Δ r H 101 =-47±9 kJ·mol -1 , Δ r S 101 =179±36 J·mol -1 ·K -1 , Δ r G 111 =-117.2±0.3 kJ·mol -1 , Δ r H 111 =-23±5 kJ·mol -1 , and Δ r S 111 =315±63 J·mol -1 ·K -1 (0.50 m NaNO 3 , 25.0±0.1°C) were determined by van’t Hoff analysis. The formation of each Tc(IV)–EDTA complex is exothermic and present favorable entropy terms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Draft Operating Procedures for a Generic Repository Licensing Organization

This document is milestone M4FT-15SN0826021: Draft Summary of principle operating procedures for a generic repository licensing organization, the result of efforts under work package FT-15SN082602 DOE Managed HLW and SNF Research: Generic Operating Procedures; e.g., Quality Assurance — Sandia National Laboratories (SNL). The objective of this effort was to develop this draft document identifying and summarizing the principle operating procedures for a generic repository licensing organization. The FY16 follow on work package is FT-165N050303: Develop and Implement Operating Procedures — SNL, having a consistent objective to develop a draft document identifying and summarizing the principle operating procedures for a generic repository licensing organization. This document identifies and describes the principle operating procedures for a hypothetical and generic organization with the responsibility to site, characterize, design, license, construct and operate a repository for the disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. These principle operating procedures reflect practices that facilitate compliance with U.S. Nuclear Regulatory Commission (NRC) expectations and NQA-1 quality standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Draft Infrastructure Framework for a Generic Repository Licensing Organization

This document is milestone M4FT-15SN0826011: Draft Summary of generic repository licensing infrastructure, the result of efforts under work Package FT-15SN082601 DOE Managed HLW and SNF Research: Establish a generic organization framework for records management, IT infrastructure, quality assurance, and systems engineering - Sandia National Laboratories (SNL). The objective of this effort was to develop this draft document identifying and summarizing the principle elements of a generic repository licensing infrastructure organizational framework. The FY16 follow on work package is FT-16SN050302- Establish organizational framework to meet regulator expectations - SNL, having the consistent objective to 'Develop a draft document identifying and summarizing the principle elements of a generic repository licensing infrastructure and organizational framework.' This document describes the framework for a generic hypothetical organization with the responsibility to site, characterize, design, license construct and operate a repository for the disposal of high-level radioactive waste (HLW) and spent nuclear fuel ( F) managed by the Department of Energy. organizational framework suggested is intended to be agnostic of the facility location and disposal medium. It is assumed that the facility is to be licensed under U.S. Nuclear Regulatory Commission (NRC) regulation with DOE as the applicant (licensee). Detailed organizational elements reflect a workforce composition and practices that facilitate compliance with NRC expectations and NQA-1 quality standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Prediction of Long-Term Geochemical Change in Bentonite Based on the Interpretative THMC Model of the FEBEX In Situ Test

Since nuclear energy is crucial in the decarbonization of the energy supply, one hurdle to remove is the handling of high-level radioactive waste (HLW). Disposal of HLW in a deep geological repository has long been deemed a viable permanent option. In the design of a deep geological repository, compacted bentonite is the most commonly proposed buffer material. Predicting the long-term chemical evolution in bentonite, which is important for the safety assessment of a repository, has been challenging because of the complex coupled processes. Models for large-scale tests and predictions based on such models have been some of the best practices for such purposes. An 18-year-long in situ test with two dismantling events provided a unique set of chemical data that allowed for studying chemical changes in bentonite. In this paper, we first developed coupled thermal, hydrological, mechanical, and chemical (THMC) models to interpret the geochemical data collected in the in situ test and then extended the THMC model to 200 years to make long-term prediction of the geochemical evolution of bentonite. The interpretive coupled THMC model shows that the geochemical profiles were strongly affected by THM processes such as evaporation/condensation, porosity change caused by swelling, permeability change, and the shape of concentration profiles for major cations were largely controlled by transport processes, but concentration levels were regulated by chemical reactions, and the profiles of some species such as pH, bicarbonate, and sulfate were dominated by these reactions. The long-term THMC model showed that heating prolongs the time that bentonite becomes fully saturated in the area close to the heater/canister; however, once the bentonite becomes fully saturated, high concentrations of ions in bentonite near the heater, which was observed in the field test, will disappear; illitization continues for 50 years but will not proceed further.

Zheng, Liange↗
DOE OSTI2019

Influence of Calcite on Uranium(VI) Sorption onto Montmorillonite Clay

Currently, most proposed methods for the long-term storage of nuclear waste include engineered buffer systems (EBS) in underground geologic repositories. The design of EBS typically involves a layer of geological buffer material such as bentonite. Bentonite is an ideal material for these systems, since it largely consists of the swelling smectite clay montmorillonite, which provides two types of surface sites available for the sorption of radionuclides, such as uranium(VI) (U(VI)). However, the presence of calcite, a common accessory mineral in bentonite, may potentially alter the sorption of U(VI) onto montmorillonite in multiple ways. First, calcite provides an additional surface for U(VI) sorption reactions. In addition, calcite dissolution will lead to dissolved calcium in porewater solutions, which may in turn affect U(VI) solution speciation, U(VI) ion exchange reactions, and the stacking and interlayer spacing of smectite layers. Last, a potential formation of U(VI)-Ca-carbonate co-precipitates may remove U(VI) from solution. As a first step, we characterized the sorption of U(VI) onto montmorillonite in the presence and absence of dissolved calcium at various ionic strength conditions in batch sorption envelope experiments. The presence of dissolved calcium resulted in decreased U(VI) sorption onto montmorillonite in acidic environments, but in no apparent changes in U(VI) sorption in near-neutral pH ranges. At higher pH values, there may be a possible decrease in U(VI) sorption, which needs to be further investigated. The observed trends at low pH can be explained by the preferential binding of Ca 2+ cations relative to UO 2 2+ , which resulted in a competition between the two cations for clay surface sites. This hypothesis is further supported by the results of additional sorption experiments, which were performed at various ionic strengths, but in the absence of calcium. In a next step, we will expand our experiments to include mixtures of calcite and montmorillonite as solids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the UNFSF. The framework for integration of safeguards and security into the UNFSF will include 1) identification of applicable regulatory requirements, 2) selection of a common system that share dual safeguard and security functions, 3) development of functional design criteria and design requirements for the selected system, 4) identification and integration of the dual safeguards and security design requirements, and 5) assessment of the integration and potential benefit.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Heat‐Generating Nuclear Waste in Salt: Field Testing and Simulation

Core Ideas A field‐scale experiment and numerical simulations confirm salt backfill behavior. Simulations closely match temperature around and under the piled salt backfill. Results indicate limited dissolution–precipitation reactions around the heat source. Alteration of backfill is unlikely if the drift is allowed to dry before emplacement. Investigations relating to in‐drift disposal of heat‐generating nuclear waste in salt have raised questions about heat–brine interactions in the unsaturated run‐of‐mine (RoM) salt pile used as backfill. These interactions have the potential to change the structure of the RoM salt surrounding the canister, possibly altering long‐term containment of the source. An experiment is in progress at the Waste Isolation Pilot Plant (WIPP), New Mexico, in which a heated canister was placed on the floor of an open drift, covered in a pile of RoM salt, and energized with 1000 W. Temperature in the RoM salt pile had stabilized after about 15 d, allowing evaluation of the heat‐up period of the ongoing experiment. Using a multiphase porous flow simulator that has been modified to handle salt‐specific coupled processes, we examined coupled thermal–hydrological–chemical behavior in the RoM salt pile. Our simulations suggest that for the relatively dry cases examined, porosity changes within RoM salt in a generic salt repository are likely to be minor in the period between waste emplacement and plastic closure of the drift. The primary sensitivity for porosity change is to the early moisture content of the RoM salt used to cover the canister. Secondary influences include moisture availability from the disturbed rock zone (DRZ) surrounding the drift and the capillary pressure ratio between the DRZ and the RoM salt. Early changes in porosity and permeability may be affected by moisture content, but this was not observed in the test. Such changes would be most likely to occur when using damp RoM salt or if waste is emplaced in a drift immediately following opening of the drift before evaporative dewatering of the drift walls occurs.

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

Radiation Chemistry Research for Improved Nuclear Fuel Cycles

Nuclear energy is a key component of a global, low-carbon energy future. One of the challenges facing large-scale deployment of nuclear power is the long-term disposition of used nuclear fuel. Due to the political and technical challenges of siting deep geologic storage facilities, reprocessing of used nuclear fuel is desirable to reduce the total volume and decay heat of waste requiring long-term storage, enabling efficient usage of geologic storage repositories. However nuclear fuel reprocessing is complicated by intense multi-component radiation fields that degrade solvent systems that are used for this purpose. Resulting radiolytic degradation can result in reduced performance over time, and generation of additional waste. Thus, an understanding of the radiation-induced chemistry of these reprocessing solvent systems is crucial to for the development of new, radiation-resistant processes for efficiently treating used nuclear fuel. Controlled irradiation experiments combined with high-performance chromatography and mass spectrometry have provided comprehensive knowledge of radiolytic degradation pathways and kinetics, which is needed to predict the effect of radiation-induced chemistry on the performance of the nuclear fuel reprocessing technology. This talk will give an overview of recent progress in research on the radiation chemistry of fuel reprocessing molecules conducted by the Idaho National Laboratory Center for Radiation Chemistry Research (CR2) in conjunction with the Molecular Mass Spectrometry and Chromatography group.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Building a Credible and Updated WIPP Performance Assessment - New Knowledge Versus Conservative Assumptions and FEP Analysis, A Balancing Act - 16299

Conservative assumptions are important components in conceptual model development where limited data is available. And, as appropriate, during the early 1990's Department of Energy (DOE) adopted many conservative assumptions in the development of the Waste Isolation Pilot Plant's (WIPP's) 24 Conceptual Models incorporated in DOE's 1996 Compliance Certification Application (CCA) performance assessment (PA). Many of these assumptions were formulated and documented in the WIPP relevant features, events and processes (FEPs) analyses; others were adopted when developing input parameter values and their distributions that were used as inputs to computer models. The Agency accepted these conservative assumptions during their 1996 CCA assessment for several reasons; 1) at the time there was a scarcity of data related to long-term behavior of many parameters needed to be incorporated in the PA, 2) sensitivity studies performed during this time period determined that the affected 'conservative' assumptions adopted for some parameters had minimal to no impact on repository performance; and, 3) DOE could demonstrate compliance with EPA's radioactive waste disposal regulations while adopting conservative assumptions and data. Since the initial 1996 CCA DOE has modified the repository panel closure design, and the repository footprint in the experimental areas has expanded. These modifications were incorporated into DOE's 2014 Compliance Re-certification Application for WIPP. Additionally, there have been numerous investigations related to both the short- and long-term behavior of salt in a repository setting. As a result, the Agency is reassessing the use of these conservative assumptions and the continued applicability of the original FEP analyses, primarily developed during DOE's 1996 PA and carried forward to the current re-certification. This paper examines some of these analyses. (authors)

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

Structure and thermodynamics of calcium rare earth silicate oxyapatites, Ca 2 RE 8 (SiO 4 ) 6 O 2 (RE = Pr, Tb, Ho, Tm)

Calcium rare earth silicate oxyapatites, (Ca 2 RE 8 (SiO 4 ) 6 O 2 ), are of interest as components of glass-ceramic nuclear waste forms. To assess their long-term behavior in a geologic repository, it is essential to determine their structure and thermodynamic stability at relevant conditions. Here, in this work, we performed detailed structural and thermodynamic investigations on Ca 2 Pr 8 (SiO 4 ) 6 O 2 , Ca 2 Tb 8 (SiO 4 ) 6 O 2 , Ca 2 Ho 8 (SiO 4 ) 6 O 2 , and Ca 2 Tm 8 (SiO 4 ) 6 O 2 by high energy synchrotron powder X-ray diffraction combined with Rietveld analysis and high temperature oxide melt drop solution calorimetry. Enthalpies of formation from constituent oxides (ΔH f,ox ) were determined to be -765.1 ± 22.8 kJ/mol for Ca 2 Pr 8 (SiO 4 ) 6 O 2 ; -638.9 ± 20.5 kJ/mol for Ca 2 Tb 8 (SiO 4 ) 6 O 2 ; -643.3 ± 10.3 kJ/mol for Ca 2 Ho 8 (SiO 4 ) 6 O 2 ; and -403.2 ± 5.1 kJ/mol for Ca 2 Tm 8 (SiO 4 ) 6 O 2 . These thermodynamic parameters were used in assessing the thermochemical stability of these phases in the presence of water vapor from room temperature to 600 K, as encountered in the subsurface environments of a geological repository.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Surrogate Model Development of Spent Fuel Degradation for Repository Performance Assessment

In model simulations of deep geologic repositories, UO 2 fuel matrix degradation typically begins as soon as the waste package breaches and groundwater contacts the fuel surface. The initial degradation rate depends on the timing of these events, burnup of the fuel, temperature, and concentrations of dissolved reactants. Estimating the initial rate of degradation is fairly straightforward, but as UO 2 corrosion products precipitate on the fuel surface and the movement of dissolved species between the fuel surface and environment is impeded by the precipitated solids, the rate is more difficult to quantify. At that point, calculating the degradation rate becomes a reactive-transport problem in which a large number of equations must be solved by iteration for a large number of grid cells at each time step. The consequence is that repository simulations, which are already expensive, become much more expensive, especially when hundreds or thousands of waste packages breach. The Fuel Matrix Degradation (FMD) model is the process model of the Spent Fuel and Waste Science and Technology (SFWST) campaign of the US Department of Energy (DOE). It calculates spent fuel degradation rates as a function of radiolysis, redox reactions, electrochemical reactions, alteration layer growth, and diffusion of reactants through the alteration layer. Like other similar fuel degradation process models, it is a complicated model requiring a large number of calculations and iterations at each time step.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Disposal of High-Level Nuclear Waste in Deep Horizontal Drillholes

Spent nuclear fuel and high-level radioactive waste can be disposed in deep horizontal drillholes in sedimentary, metamorphic or igneous rocks. Horizontal drillhole disposal has safety, operational and economic benefits: the repository is deep in the brine-saturated zone far below aquifers in a reducing environment of formations that can be shown to have been isolated from the surface for exceedingly long times; its depth provides safety against inadvertent intrusion, earthquakes and near-surface perturbations; it can be placed close to the reactors and interim storage facilities, minimizing transportation; disposal costs per ton of waste can be kept substantially lower than for mined repositories by its smaller size, reduced infrastructure needs and staged implementation; and, if desired, the waste could be retrieved using “fishing” technology. In the proposed disposal concept, corrosion-resistant canisters containing unmodified fuel assemblies from commercial reactors would be placed end-to-end in up to 50 cm diameter horizontal drillholes, a configuration that reduces mechanical stresses and keeps the temperatures below the boiling point of the brine. Other high-level wastes, such as capsules containing 137 Cs and 90 Sr, can be disposed in small-diameter horizontal drillholes. We provide an overview of this novel disposal concept and its technology, discuss some of its safety aspects and compare it to mined repositories and the deep vertical borehole disposal concept.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The effect of cerium, neodymium, and ytterbium doping on UO 2 dissolution

Here, the dissolution rate of spent nuclear fuel has been studied heavily to understand the impacts of a failed waste package scenario on potential radionuclide release from a geologic repository. Multiple countries have evaluated these scenarios in oxidizing and reducing environments respective to their relevant repository conditions. The repository environment (e.g., water chemistry, temperature, oxygen content) and the fuel itself (e.g., chemical content of the fuel) both heavily impact the dissolution rate of the spent fuel. This work examined the impact that rare earth element dopants (Ce, Nd, Yb) have on the fuel dissolution under repository relevant conditions with decreasing oxidizing conditions using a single pass flowthrough system. UO 2 samples were doped with Ce, Nd, or Yb with concentrations between 1 and 5 at%. The addition of dopants to the samples reduced the dissolution rates on most samples relative to pure UO 2 samples. Scoping experiments that occurred in a less oxidizing environment showed a reduction in dissolution rate compared to fully oxidizing conditions. The results within highlight the importance of dopant behavior in used fuel dissolution modeling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Draft Experimental Plan for Commercial SNF Degradation in Repository Environments with a Focus on Fracture Matrix Degradation

This report documents work performed supporting US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.04, “Inventory and Waste Form Characteristics and Performance.” In particular, this report fulfills the M3 milestone, M3SF-21OR010309072 "Draft Experimental Plan for Commercial SNF Degradation in Repository Environments” as described in work package SF-21OR01030907 "SNF Degradation Testing-ORNL".This experimental plan is a collaboration between ORNL, SNL, ANL, and PNNL. It is intended to support source term model development and spent fuel degradation testing in a range of generic repository conditions. Potential testing and experiments to support validation of the Fuel Matrix Dissolution Model (FMDM) and the overall development/validation of repository source term modeling are discussed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗