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DOE OSTI2015

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

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

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

Brine Availability Test in Salt (BATS) (FY25 Final Closeout)

This fiscal year 2025 (FY25) closeout report summarizes the final experiments conducted and data collected in phase 2 of the Brine Availability Test in Salt (BATS 2). The design and interpretation of the test was previously funded by the US Department of Energy’s Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition Program, under the Disposal Research and Development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) but they are not funding it anymore (since May 2025). The experiment was located underground at the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico, which is a DOE Office of Environmental Management (DOE-EM) site managed by the Carlsbad Field Office (CBFO). DOE-EM funds the WIPP Test Coordination Office (TCO), which provided critical implementation support for the continued execution of BATS. A high-level summary of processes expected in a salt repository are given in Kuhlman & Mills (2025).

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

Impacts of Lanthanide Ion Complexation on the Radiation Robustness of Diglycolamide Extractants

Nuclear power reactors represent a clean and reliable source of baseload energy. However, nuclear reactors produce used nuclear fuel (UNF), which, if not recycled/reprocessed, must be disposed of as high-level radioactive waste, necessitating long-term storage options. UNF contains approximately a third of the periodic table, including the radioactive minor actinides (MA) americium and curium. Removal of these MA from UNF would greatly reduce the radiotoxic burden, volume, and cost of a storage facility or geological repository. Thus, a significant global effort has been devoted to the design of reprocessing strategies for the clean separation of the MA from UNF. As MA separations are difficult, due to the presence of lanthanide elements in UNF and their remarkably similar chemical properties, many extractants have been studied. The diglycolamide (DGA) class of extractants are promising for this separation as they exhibit high distribution coefficients for the MAs, high metal loading capacities, and are robust in highly acidic radiation environments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Compressible Arch Segments for the Cigeo Access Ramps, Drifts and Vaults - A Field Test - 17011

Andra is currently implementing the detailed engineering phase of the step-wise design development of Cigeo (the French HLW and IL-LLW Deep Geological Repository). Various technical issues are at stake, including the requirements linked to 'Reversibility / Retrievability', at different stages of the repository life cycle (from its construction start-up to its final closure and monitoring). These reversibility requirements have a significant impact on the way Cigeo is designed, will be progressively developed, operated, later closed and monitored. On a shorter term, one must demonstrate that the waste container retrieval operations are technically safe and feasible. In that sense, the structural dimensioning of the ramps and mains, as well as that of the IL-LLW disposal vaults, is critical, since the concrete liner supporting the rock walls must last over a period of some 100 to 150 years (lapse of time during which retrieval operations are likely to take place). The present paper starts with a description of the geomechanical conditions prevailing in the Callovo-Oxfordian (COX) clay-stone formation (aka the argillite). The structural dimensioning of the IL-LLW disposal vault liner, as calculated at the end of the basic engineering phase is indicated. The thickness values obtained so far have led to considering an alternative design in which a compressible material is positioned between the concrete liner intrados and the rock walls: in such situation, the 'creeping effect' of the rock is mitigated, for a time, by the compressible material (nearly a perfect elastoplastic material), hence reducing the stress level exerted inside the concrete liner. This situation enables to significantly reduce the liner thickness and provides potentially significant saving in the construction of the works concerned (preliminary sensitivity results are given). The following part of the paper is then focused on a practical case story: a technological demonstration (still ongoing) of a tunnel boring experiment, implemented underground (in a dedicated drift of the Meuse Haute Marne URL infrastructures, at scale 1:2 of the Cigeo drifts/vaults). In this excavation and construction experiment, the tunnel boring machine (TBM) lays (via an erector) 'classical' concrete arch segments (forming a ring) in a first portion of the drift and 'compressible' ones (the compressible material is positioned on their extrados) in a second portion of the drift. The technical demonstration objective is two-fold: the capacity to lay compressible arch segments with a TBM is checked, their structural behavior (with time) is compared with that of 'classical' ones. The way the compressible material is produced and then fixed on the arch concrete segment is also documented, c/w a description of the complementary testing campaign to be carried out on a test bench (on surface) of a concrete ring (at a scale 1:2 of Cigeo) formed of compressible arch segments. This paper concludes with a timeline of the experiment. Should the results obtained provide some perspectives of technical and economic improvement (i.e. saving vis-a-vis the 'classical' solution) a decision could be taken by Andra to incorporate this alternative solution in the 'design freeze' of Cigeo underground infrastructures. The positive outcomes of this technical test campaign would also pave the way for specifying the Cigeo TBM characteristics (dimensions and handling equipment), scheduled for manufacturing order around 2020. (authors)

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

A Study Of Geological Disposal Facility (GDF) Ageing Phenomena In Bentonite - 17605

Since the 1950's, the UK's nuclear activity in energy, defence and medicine has led to the accumulation of a large amount of higher activity radioactive waste (HAW). Most of nuclear waste arises in a wide variety of physical and chemical forms with different radioisotopes and loadings. Due to complexity in HAW, it is imperative that there be a rigorous management system that ensures population and environmental safety over extended periods. The concept of Deep Geological Disposal of waste in a stable geological formation is, nowadays, an internationally accepted solution as the surest and the most viable long term method for management. Any future Geological Disposal Facility (GDF) will comprise, among other artificial and natural engineered barriers, a barrier immediately around the waste containers compounded by a sealing material. The long-term stability of a GDF will be determined by the design and construction of the engineered barriers, and the mechanical, hydraulic and geochemical interactions that may happen between the different engineered components as a result of combined processes such as the heat produced by radioactive decay and the ion water composition of the surrounding host rock. Bentonite clay has been selected as the most promising buffer and backfill material in many disposal programmes because of its swelling capacity, low permeability, plasticity and cation exchange capacity. In the nuclear scope, bentonite testing at different scale-lengths (laboratory and field scale) has been performed and implemented in order to examine the thermo-hydro-mechanical (THM) and thermo-hydro-geochemical (THG) processes that may happen in the engineered barriers and geological settings associated with a GDF. Two major European field scale tests have been performed to better understand bentonite performance and behaviour immediately after HAW emplacements. The Full-Scale Engineered Barriers Experiment (FEBEX) at the Grimsel Test Sit (Switzerland) and the Alternative Buffer Material (ABM) Project at the Aespoe Hard Rock Laboratory (Sweden). The Febex Experiment was a pilot project designed for 'field research' of HAW repository hosted in crystalline rock where surrogate HAW canisters were placed horizontally in drifts, surrounded by compacted blocks of bentonite, sealed by a concrete plug and maintaining a constant canister surface temperature of 100 deg. C. After five years of operation, a partial dismantling was carried out in 2002. Heater processes occurring in the surrounding bentonite. The dismantling of the heater no. 2 (after 18 years of operation) started in late 2014 and is planned to be finished at the end of 2016 (termed FEBEX-DP). The FEBEX Project utilised bentonite from the Cortijo de Archidona deposit (Almeria, Spain). This 'Febex' bentonite is sourced from calcium-magnesium montmorillonite with good properties in relation to swelling capacity, permeability, adsorption capacity, and thermal conductivity. The ABM Project is a large-scale programme that has been running since 2006. The programme is designed to study a suite of possible buffer materials and correlate the physical and chemical properties of eleven different types of bentonites to their fundamental mineralogical properties. The ABM test consists in three 'packages' containing a central heater tube surrounded by different pre-compacted bentonite blocks and instrumentation. Each 'package' was installed in separate boreholes. The target temperature for the surface of the heated tubes in contact with the bentonite was 130 deg. C. Both FEBEX and ABM test the bentonite was saturated with water percolating from surrounding host rock (FEBEX and ABM) and also via artificial supply systems (ABM), being the water itself acting as a long term corrosion agent. For both experiments and emerging interest is in the interaction of bentonite with metallic components, under geological disposal conditions are not considered to be thermodynamically stable and hence arising corrosion products could interact with the buffer. These interactions depend on various factors such as the redox potential, temperature, pH, solute concentration and buffer material characteristics. During the operational phases of the FEBEX and ABM and after its 'closure' a significant volume of air was found to be retained in the bentonite. Under these conditions a period of initial aerobic corrosion by the air in the bentonite barrier would have occurred followed by establishment of anaerobic corrosion would be the dominant process ongoing. The anaerobic corrosion began when the bentonite became water saturated. Under these conditions, the water acted as the major oxidant. The present project focuses on the characterization of samples arising from the FEBEX and ABM Projects as well as performing additional independent experiments designed to complement these prior studies. Characterization will focus mainly on corrosion related alteration combined with thermal effects. Through detailed sample characterization, the intention is to achieve a greater knowledge regarding the changes to the bentonite under the experimental conditions to which it was exposed, and plausible alterations in its THM and THG properties occurring during the experiment due to blended effects of temperature,water content, engineering structural components and solutes are also to be measured. The fulfilment of a large-scale 'in-situ' test is intricate, time-consuming and with an abundance of external variables that are not always well controlled. For this reason, the University of Bristol is performing a series of well constrained time-resolved laboratory tests which bentonite is being subjected to different possible scenarios that may happen whilst in a HAW repository. In virtue of this experiment, the UoB aims to understand the illitisation and cementation process and their effects on bentonite safety properties. (authors)

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

An Evaluation of Standardized Canisters in the Waste Management System - 16317

Development of a standardized canister system represents an opportunity to develop an integrated approach to addressing storage, transportation, and disposal issues in the waste management system. However, regardless of timing and method, deployment of such a system would have the potential to cause significant system-wide impacts. This evaluation expands upon previous standardized-canister system evaluations and compares continued loading of dual-purpose canisters (i.e., the status quo) with loading of standardized canister systems in the near term, that is, before repository requirements are known and/or before operating reactors shut down. This evaluation quantitatively compares order-of-magnitude costs and logistics for different standardization scenarios with those of status quo scenarios, provides insight into quantifiable impacts of loading standardized canister systems in the near term, tests system-level analysis tools and associated input, and identifies scenarios for further analysis. This is a technical paper that does not take into account the contractual limitations under the Standard Contract (10 CFR Part 961). As work on assessing standardized canister system scenarios has progressed, new data for at-reactor and interim storage facility (ISF) operations have been collected and/or generated to provide more realism at the system level, and past scenarios were re-evaluated with this new information. In addition, new scenarios were developed and analyzed that focused on (1) bare fuel transported from reactors to an ISF in re-usable bare fuel casks and deferring deployment of a standardized canister system to the ISF, (2) alternative acceptance strategies based on conclusions from previous systems architecture studies, and (3) the impacts of having to accommodate spent fuel in differently sized canisters for different potential repository geologies. Results of this evaluation indicate that shifting the loading of standardized canisters from the reactor to the ISF does not result in a significant change in total system cost. Results also show that some alternative acceptance strategies could reduce the number of years that spent nuclear fuel (SNF) stays on-site at reactors for cases involving standardized canisters (consistent with previous results that looked at dual-purpose canisters). Finally, incorporating the new at-reactor loading data of small standardized canisters shows a significant system-wide cost reduction when compared to previous estimates of at-reactor loading of small canisters. (authors)

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

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

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

Reduced-order condensed-phase kinetic models for polyethylene, polypropylene and polystyrene thermochemical recycling

Thermochemical recycling of plastic waste (PW) into chemicals and energy vectors requires coupling particle and reactor-scale simulations to accurate condensed phase pyrolysis mechanisms for each constituent. This work proposes a methodology to derive reduced-order condensed-phase kinetic models from validated semi-detailed kinetic mechanisms. Two types of kinetic models are obtained for polyethylene (PE), polypropylene (PP) and polystyrene (PS): reduced semi-detailed models and multi-step fully lumped ones. These families offer different compromises between accuracy and computational cost. The former employ 50–100 gas + liquid species and describe both the radical degradation and the detailed carbon distribution of the products. Conversely, the latter involves 5–10 species per polymer tracking only the main petroleum cuts. The kinetic mechanisms are complemented by the definition of thermochemical properties of gas, liquid, and solid-phase species, accounting for phase-transitions through pseudo-chemical reactions. Model validations are performed by comparison with experimental data and the original semi-detailed mechanisms in terms of mass loss, heat fluxes and product distribution profiles. The resulting CHEMKIN-like condensed-phase models are attached as Supplementary Material and as a GitHub repository. Extending the proposed approach to other polymers and coupling it with existing subsets in the CRECK kinetic framework (e.g., biomass, PVC, PET) offers a powerful tool to model thermochemical recycling of PW and biomass/PW mixtures.

kinetics↗
DOE OSTI2016

Perspective of Dry Interim Storage of spent nuclear fuel (SNF) in Germany - 16504

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection (BfS). In 1995 a license to store spent nuclear fuel (SNF) and vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2035. During the last two years in Germany preparatory consultations to restart the process towards finding a repository took place, organised by a commission. This commission consists out of members of the scientific community, religious parties, social communities and environmental organizations as well as members of federal states governments and members of the German parliament (Bundestag). Recently the commission decided to extend its working period until summer 2016. Taking this background into account the BfS plans to intensify its research into extended interim storage with respect to potential material degradation leading to interference with safety functions (e.g. neutron shielding material). This includes topics like ageing of gasket materials which are already studied in ongoing experiments (e.g. at the Federal Institute for Materials Research and Testing) as well as studies of long term relaxation processes. Because of the need for transportation towards a repository another important field is the behaviour of the inventory under mechanical loads. For SNF, first calculations showed that because of the lower decay heat load after 60 years due to hydride reorientation the cladding might become much less ductile. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Inventory of Shale Formations in the US, Including Geologic, Hydrological, and Mechanical Characteristics

The objective of this report is to build upon previous compilations of shale formations within many of the major sedimentary basins in the US by developing GIS data delineating isopach and structural depth maps for many of these units. These data are being incorporated into the LANL digital GIS database being developed for determining host rock distribution and depth/thickness parameters consistent with repository design. Methods were developed to assess hydrological and geomechanical properties and conditions for shale formations based on sonic velocity measurements.

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

Interaction of Cesium and Barium Ions with Calcium Silicate Hydrate under a Geological Temperature Condition - 16239

The interaction of calcium-silicate-hydrate (CSH gel) and cationic radionuclides such as cesium-135 significantly affects the assessment of radionuclide migration. This study examined the influence of temperature on the interaction of Cs and barium (Ba) ions with CSH gel because temperature generally affects chemical reactions such as sorption. In the sorption experiments, CSH gel was prepared by the hydration of CaO and fumed silica without a drying process, with Ca/Si molar ratios ranging from 0.4 to 1.6 and the liquid/solid weight ratio set to 20. The concentration of Cs and/or Ba ions was adjusted to 1.0 mM. The temperature conditions during the sorption experiments for 14 days were set to 298, 313 and 333 K. As a result, the sorption of Cs ions on CSH gel increased at higher temperature. Moreover, Raman spectroscopy studies showed that the silicate chain of CSH gel depolymerized with the increase in temperature. That is, the structural deformation of CSH gel at a relatively high temperature facilitates the sorption of Cs ions. On the other hand, the sorption of Ba ions on CSH gel remained high (90 %) regardless of temperature, and hardly affected the sorption of Cs ions. These findings suggest that CSH gel around the repository is expected to further retard the migration of cationic radionuclides under temperature condition up to 333 K. (authors)

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

Study on Characterization of Colloidal Silica Grout for Rock Excavation Under Saline Groundwater - 19332

A rock excavation grouting technology has been recently studied as significant technology for reducing the ingress of water into the deep repository. However, it has not been studied for applying to the coastal region, where it is discussed to be a more suitable region for the geological disposal in Japan. The latest material called colloidal silica grout (CSG) is good for sealing narrow fractures but is known to be sensitive to the salinity of groundwater because of its gelling property with salt accelerator. Although the gelling of CSG can be controlled by adding an acidic pH adjuster, the methodology for delivering the appropriate grout is not well established for such conditions of saline groundwater. Therefore, this research project was established to enhance the existing rock grouting technology for deep repositories. Initially, a grouting workshop was held with Nordic authorities in order to understand the state of the art and identify the most significant challenges to developing a generic mixing method for injecting CSG under seawater. Based on laboratory tests, we developed a feasible mixing methodology that prevents the rapid reaction with the salty water, which shows that CSG is not negatively impacted by salty mixing water or submerged by saline water. Combining the latest grouting techniques from Japan and the Nordic countries creates an optimal grouting methodology for use under seawater. Using this hybrid approach, the CSG can find wide-scale use as an agent for grouting narrow rock fractures filled with saline or fossil water in order to decrease the ingress of water in the disposal facilities. The reduction of the water ingress will enhance the safety of the repository, reduce water discharge, and diminish the long-term cost of water treatment. (authors)

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

Developing reliable safeguards seals for application verification and removal by State operators

Once a geological repository has begun operations, the encapsulation and disposal of spent fuel will be performed as a continuous, industrial-scale series of processes, during which time safeguards seals will be applied to transportation casks before shipment from an encapsulation plant, and then verified and removed following receipt at the repository. These operations will occur approximately daily during several decades of Sweden's repository operation; however, requiring safeguards inspectors to perform the application, verification, and removal of every seal would be an onerous burden on International Atomic Energy Agency's (IAEA's) resources. Current IAEA practice includes allowing operators to either apply seals or remove them, but not both, so the daily task of either applying or verifying and removing would still require continuous presence of IAEA inspectors at one site at least. Of special importance is the inability to re-verify cask or canisters from which seals have been removed and the canisters emplaced underground. Successfully designing seals that can be applied, verified and removed by an operator with IAEA approval could impact more than repository shipments, but other applications as well, potentially reducing inspector burdens for a wide range of such duties.

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

Experimental determination of lead carbonate solubility at high ionic strengths: A Pitzer model description

In this article, solubility measurements of lead carbonate, PbCO 3 (cr), cerussite, as a function of total ionic strengths are conducted in the mixtures of NaCl and NaHCO 3 up to I = 1.2 mol•kg –1 and in the mixtures of NaHCO 3 and Na 2 CO 3 up to I = 5.2 mol•kg –1 , at room temperature (22.5 ± 0.5 °C). The solubility constant (log K sp ) for cerussite, PbCO 3 (cr) = Pb 2+ + CO 3 2- was determined as –13.76 ± 0.15 (2σ) with a set of Pitzer parameters describing the specific interactions of PbCO 3 (aq), Pb(CO 3 ) 2 2- , and Pb(CO 3 )Cl – with the bulk-supporting electrolytes, based on the Pitzer model. The model developed in this work can reproduce the experimental results including model-independent solubility values from the literature over a wide range of ionic strengths with satisfactory accuracy. The model is expected to find applications in numerous fields, including the accurate description of chemical behavior of lead in geological repositories, the modeling of formation of oxidized Pb–Zn ore deposits, and the environmental remediation of lead contamination.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

The underground research facilities (URF) network - illustrating the IAEA's concept and use of professional networks - 15382

Since 2001, the IAEA has championed the concept and use of professional networks (Communities of Practice) to advance best practices and to provide a platform for cooperation in radioactive waste management, decommissioning and environmental remediation across the globe. The Underground Research Facilities (URF) Network was the first in a series of networks implemented under the auspices of the IAEA to share experiences, elicit best practices, and transfer knowledge on topics relevant to the development and implementation of geological disposal of long lived and high level radioactive waste - i.e. intermediate level waste (ILW), high-level waste (HLW) and/or spent nuclear fuel (SNF). Special emphasis is placed on the role and use of URFs to support such geological disposal developments. In addition to the direct benefits provided to Member States through its programme of work, the URF Network also delivers less tangible, yet equally valuable benefits through the informal exchange of information, the development of peer-to-peer professional networks, and seeding of multi-lateral cooperation between Member States. Network members represent a range of nuclear programme sizes and advancements in their respective repository programmes. After a decade of fruitful cooperation, delivering more than 30 training courses on topics relevant to geological disposal, the URF Network is gradually reassessing its mode of operation and focus of work. Emphasizing selected strategic goals - focused in part on more systematic training of URF Network members human resources and in part on select key technical issues - would benefit all network members, while remaining cautious in the commitment of scarce resources. (authors)

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

Salt International Collaborations (FY22 Update)

This report summarizes the international collaborations conducted by Sandia funded by the US Department of Energy Office (DOE) of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D and Salt International work packages. This report satisfies the level-three milestone M3SF-22SN010303063. Several stand-alone sections make up this summary report, each completed by the participants. The sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS), granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), numerical model comparison (DECOVALEX) and an NEA Salt Club working group on the development of scenarios as part of the performance assessment development process. Finally, we summarize events related to the US/German Workshop on Repository Research, Design and Operations. The work summarized in this annual update has occurred during the COVID-19 pandemic, and little international or domestic travel has occurred. Most of the collaborations have been conducted via email or as virtual meetings, but a slow return to travel and in-person meetings has begun.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Modeling Radionuclide Releases from TRISO Particles by Simultaneous Diffusion Through and Corrosion of the Silicon Carbide Barrier Layer

TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Numerical evaluation of thermal effects from nuclear waste disposed in horizontal drillholes

© 2019 American Nuclear Society, Inc. We examine the temperature evolution near horizontal drillholes designed for the disposal of spent nuclear fuel and high-level radioactive waste. In this concept, directional drilling technology is used to complete a small-diameter, vertical access hole, which at the target depth gradually turns into a horizontal section dedicated for the end-to-end disposal of nuclear waste canisters in a suitable, low-permeable formation. Decay heat emanating from the waste leads to temperature increases within the drillhole and the surrounding host rock. Predicting the thermal evolution within the engineered and natural barrier systems is important, because high temperatures may lead to thermal stresses, alter material properties, or induce driving forces that affect the migration of radionuclides along the drillhole and in the near field of the repository. We present the results of numerical design calculations that examine the spatial and temporal evolution of temperature as a function of uncertain material properties and adjustable design parameters.

Finsterle, S↗