Focused Session IC-5.H
Materials Issues in Radioactive Nuclear Waste Treatment and Disposal
ABSTRACTS
Session IC-5.H.1 Waste form development, including glass, ceramic, cement and metallic waste forms
IC-5.H.1:IL01 Glass-ceramic Waste Forms for Actinide Immobilization
D.J. GREGG, Australian Nuclear Science and Technology Organisation, Kirrawee DC, NSW 2232, Australia
Actinide-bearing waste streams that arise from the nuclear fuel cycle present unique scientific and technical challenges. These wastes can range from pure to very impure forms, have very long half-lives and can contain fissile material (e.g., Pu-239), thus requiring criticality control and safeguards measures. The immobilisation of this material requires significant interconnection between the wasteform and the process technology to enable production of a suitably durable product, whilst maintaining flexibility to accommodate waste variations and satisfy criticality prevention requirements. Synroc-type titanate phases have long been associated with the treatment of actinide wastes partly due to their demonstrated ability to immobilise actinides in the geological environment for billions of years. The extension to glass-ceramic advanced wasteforms combines the advantages of conventional borosilicate nuclear waste glasses and their chemical and processing flexibilities with the superior chemical durability of the ceramic phases which host the actinides. This paper will provide an overview of fundamental research activities at ANSTO focusing on glass-ceramic wasteform design, fabrication and processing, chemical durability and ion-beam irradiation studies.
IC-5.H.1:IL02 Progress in Tuning Cementitious Materials for Enhancing Retention of Radionuclide Anions
R.M. ASMUSSEN, Pacific Northwest National Laboratory, Richland, WA, USA
Cementitious materials are critical to nuclear waste management strategies worldwide. Their formulations must be tailored to specific waste types and disposal environments to ensure long-term containment. A key performance metric is the ability to retain contaminants and radionuclides over the operational lifespan of the disposal facility. While conventional cementitious systems effectively immobilize many species, certain long-lived radionuclides (e.g., Tc-99, I-129) and contaminants (e.g., Cr) exhibit limited retention under standard conditions. Recent research has focused on enhancing retention through formulation tuning and the use of targeted additives known as getters. These approaches include mineral incorporation (e.g., Tc-99 into ettringite), reductive armoring (e.g., Tc and Cr into Fe-oxyhydroxides), solubility control (e.g., I-129 precipitation with Ag or Bi), and selective sorption (e.g., anion-binding additives). This presentation will highlight current advances in designing cementitious materials that improve immobilization of these challenging species.
IC-5.H.1:IL03 Incorporation of Chlorides in Oxide Glasses and Glassy Wasteforms
P.A. BINGHAM, L. COOPER, A. SCRIMSHIRE, G. TAMENI, S. SAMAD, A.M.T. BELL, School of Engineering and Built Environment, Sheffield Hallam University, City Campus, Sheffield, UK
Chlorides, in different chemical forms and at different concentrations, occur somewhat ubiquitously throughout radioactive wastes, toxic wastes and other wastes that are destined for vitrification or thermal treatment. However, chlorides are challenging to incorporate safely within oxide glasses, with problems relating to chloride solubility and volatility requiring solutions. Wastes in which chloride is present include (i) non-radioactive slags, ashes and other wastes that could be valorised through thermal treatment; (ii) some high-level, intermediate level and low activity radioactive wastes (HLW, ILW and LAW, respectively), including decommissioning wastes, that are being (or may be) vitrified to reduce volumes and create passively-safe vitreous wasteforms; and (iii) a number of future wastes that may arise from new reactor types, new nuclear fuel cycles and future fuel pyroprocessing activities. In this presentation we will explore current understanding of chloride incorporation in oxide glasses, and the factors that govern it. We will summarise literature from a range of fields, from archaeological glasses to non-nuclear waste glasses, to examples from the nuclear arena. We will then present our recent work on developing new glass formulations that may be capable of incorporating alkali chloride-rich wastes arising from future molten salt pyroprocessing activities. The aim of this presentation will be to raise awareness of the issues surrounding chloride incorporation in oxide glasses, and to stimulate discussion and debate relating to development of new glass formulations and processing routes capable of generating increased levels of chloride incorporation in oxide glasses and vitreous wasteforms.
IC-5.H.1:IL04 New Ceramic Phases for Minor Actinides Immobilization in TiO2-ZrO2 Based Systems
CHANGZHONG LIAO, JIAQI XU, YURONG SHI, Guangxi University, Nanning, China
The management of high-level radioactive waste (HLW) demands durable matrix materials that can incorporate and securely retain a broad range of radionuclides over geological timescales. Ceramic waste forms show great promise for the immobilization of HLW. In this study, we investigated novel ceramic matrices within TiO2–ZrO2-based multi-component systems as potential candidates for waste immobilization. In the CaO–TiO2–CeO2 system, we identified and synthesized a pure Ca-deficient pyrochlore phase and conducted a comprehensive investigation of its crystal structure, solid solubility, cation oxidation states, and Raman response under high pressure. We also identified a previously unreported compound in the TiO2–ZrO2–CeO2 system that exhibits a superlattice structure derived from TiZrO4. Detailed crystallographic analysis shows mixed Ti and Zr occupancy at distinct Wyckoff sites, revealing long-range cation ordering within the lattice. The new Ca-deficient pyrochlore and TiZrO₄-derived superlattice compound are promising materials for safely immobilizing nuclear waste, as they can incorporate actinides without losing structural stability under repository conditions.
IC-5.H.1:IL05 CERMET Materials for Waste Forms
J.S. McCLOY1,2, M.C. DIXON WILKINS1, J. EVARTS1,2, A. GOEL3, R. SAINI3, S.K. SUNDARAM4, N. MARRERO4, J. AMOROSO5, J. VIENNA2, J. CRUM2, A.A. MAKSIMOVA6, H. ZUR LOYE6, T. HAN7, A. KUMAR7, 1Washington State University, Pullman, WA, USA; 2Pacific Northwest National Laboratory, Richland, WA, USA; 3Rutgers University, Piscataway, NJ, USA; 4Alfred University, Alfred, NY, USA; 5Savannah River National Laboratory, Aiken, SC, USA; 6University of South Carolina, Columbia, SC, USA; 7Missouri University of Science & Technology, Rolla, MO, USA
Cermet composites, combinations of ceramic and metallic phases, offer a high-performance route for immobilizing complex waste streams from advanced reactors, including TRISO fuel debris, pyroprocessed fluoride and chloride salts, and metallic residues. These materials integrate the chemical durability of ceramics with the toughness and thermal conductivity of metals, enabling waste loadings above 70 vol%. Model systems with 316L stainless steel (SS) matrix and SiC or ZrO2 ceramic phase when consolidated by hot uniaxial pressing or spark plasma sintering (≤1100 °C, 30–40 MPa) exhibit low porosity and strong interfacial bonding. SS-SiC cermets show formation of Fe₅Si₃, NiSi₂, and graphite, with segregation of Cr and Ni across metallic regions. Incorporation of UO₂ and oxyhalide ceramics such as sodalite and CsNb₂O₅F extends the approach to fission product and salt waste. Cermet architectures enable moderate-temperature processing, high mechanical integrity, and chemical stability, offering a flexible, scalable waste form for diverse high-level wastes from Gen IV reactor fuel cycles, including molten salt and TRISO systems. Demonstration cermets have been scaled to 50 g ceramic phase batches and 12” diameter cermets.
IC-5.H.1:IL06 The Use of Sustainable and Novel Binders to Immobilize Radioactive Wastes
QUOC TRI PHUNG, L. FREDERICKX, Waste & Disposal Expert Group, Belgian Nuclear Research Centre (SCK CEN), Mol, Belgium
The nuclear sector faces significant challenges in managing organic radioactive waste, including mineral oils, tributyl phosphate, and other nuclear fuel processing by-products. These materials pose considerable waste management challenges due to their chemical reactivity and radioactive isotopes. Traditional cementation methods face issues such as poor chemical compatibility and the inability to fully stabilise organic materials. Vitrification, although effective for some waste types, demands high-energy processing, increasing costs and environmental concerns. These challenges necessitate a novel immobilisation method that effectively stabilises organic radioactive waste while remaining cost-efficient. We have successfully developed and tested geopolymer matrices designed to immobilize both liquid and solid organic radioactive wastes. These geopolymers have been tailored to encapsulate various organic waste types, offering high chemical resistance, low transport properties, and robust structural integrity. We also assessed the long-term performance under service conditions, including resistance to radiation, thermal degradation, and environmental degradation. Our findings demonstrated that metakaolin and blast furnace slag-based geopolymers are promising matrices for organic waste immobilization, accommodating up to 40 vol.% waste loading. Depending on the waste type, these materials can encapsulate waste either physically or chemically. The waste forms exhibited strong resistance to carbonation, leaching, alkali-silica reaction, and fire. This innovation has the potential to revolutionize nuclear waste management by providing a versatile, scalable, and environmentally sustainable solution for organic radioactive waste. The geopolymer-based waste forms offer significant cost savings, while enhancing public and environmental safety thanks to their superior long-term performance.
IC-5.H.1:IL07 Dissolution Behaviour of Glass Composite Materials from Reconditioning of Belgian Radioactive Waste-Part 1: Experiments and Results
K. FERRAND, S. CAES, K. LEMMENS, P. SCHROEDERS, E. VALCKE, P. VERHEYEN, Belgian Nuclear Research Centre (SCK CEN), Mol, Belgium; E. COPPENS, K. MEERT, ONDRAF/NIRAS, Brussel, Belgium
As part of the STAB-SLAK project, which investigates the stability of slags resulting from plasma incineration-vitrification of short-lived Low and Intermediate-Level radioactive Waste in cementitious environments, leaching tests were conducted. Slags, which are glass composite materials, with varied composition were produced by heating oxide mixtures to 1400 °C for one hour, followed by fast or slow cooling. Static tests were performed at low slag surface area-to-solution volume ratio (10 m-1) to maintain a kinetically controlled dissolution regime and assess the influence of slag composition on dissolution rates. Slag grains (1–2 mm) were used to preserve microstructural heterogeneity and macro-inhomogeneity. To simulate conditions expected at the surface disposal site, a synthetic cementitious water (pH = 12.5) was used as leachant. Tests were conducted at 40 °C under anoxic conditions. Solution aliquots were regularly withdrawn, acidified, ultrafiltered, and analysed by ICP-MS to quantify element release. Results showed incongruent dissolution, and elemental release dependent on the slag composition. To categorise the leaching elements and develop a quantitative dissolution model, the leaching data were statistically processed (Coppens et al., CIMTEC2026).
IC-5.H.1:L08 Dissolution Behaviour of Glass Composite Materials from Reconditioning of Belgian Radioactive Waste-Part 2: Predictive Modelling
E. COPPENS, K. MEERT, ONDRAF/NIRAS, Brussel, Belgium; S. CAES, K. FERRAND, K. LEMMENS, P. SCHROEDERS, E. VALCKE, P. VERHEYEN, Belgian Nuclear Research Centre (SCK CEN), Mol, Belgium
Safe disposal of radioactive waste requires its compliance with acceptance criteria. For waste types failing to demonstrate this, reconditioning by plasma incineration–vitrification might be envisaged. The end product, glass composite materials (GCMs), must show chemical durability to ensure long-term safety. In the STAB-SLAK project we investigated the effect of composition and cooling rate on the dissolution behaviour of GCMs. A parametric screening was designed using a Plackett–Burman design, covering 23 experimental runs with varied oxide ratios and cooling modalities. Non-radioactive GCMs were synthesized and subjected to leach tests (Ferrand et al., CIMTEC2026). Elemental mass loss was modelled by second-order polynomials in function of time. Elements could be grouped into fast-, intermediate-, and slow-leaching clusters based on their dissolution kinetics and statistical analysis revealed the significant influence of componential factors. Mathematical models were developed enabling prediction of leach rates, supporting the translation of safety requirements into compositional constraints. These results provide a scientific basis for guiding the design and qualification of the GCMs for disposal.
IC-5.H.1:IL09 Structural Evolution in Complex Ceramics under Irradiation
CHENXU WANG1, YUGANG WANG1, QING HUANG2, JINGYANG WANG3, JIE ZHANG3, FEI GAO4, C.L. TRACY5, R.C. EWING5, 1State Key Laboratory of Nuclear Physics and Technology, Center for Applied Physics and Technology, Peking University, Beijing, China; 2Ningbo Institute of Material Technology and Engineering, Chinese Academy of Science, Ningbo, China; 3Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, Shenyang, China; 4Department of Nuclear Engineering and Radiological Science, University of Michigan, Ann Arbor, MI, USA; 5Center for International Security and Cooperation, Stanford University, Stanford, CA, USA
Diverse, multi-stage phase transformations occur in many waste forms under extreme environments. In response to irradiation, complex oxides (such as pyrochlores) undergo phase transformation to intermediate defect fluorite, some becoming amorphous at higher irradiation doses. Similarly, MAX phases possess the capability of accommodating actinide elements, indicating their potential as candidate materials for the immobilization and disposal of nuclear waste. Some MAX phases transform from an initial hexagonal structure to an intermediate γ-phase, then to a face-centered cubic (fcc) structure, while others instead become amorphous. To date, almost all analyses were based on their initial phase or structure, yet this cannot fully explain the phase transformation and amorphization behavior under irradiation. In this work, we combine in situ ion irradiation, Transmission electron microscopy (TEM), and density-functional theory (DFT) calculations to demonstrate the distinct transformation pathways in a series of MAX phases and complex oxides. We show that structural distortion and bond covalency of the intermediate phases (γ-phase and defect fluorite) determine the outcome of the transformation process. These results provide insight into the multi-stage phase transformation pathways.
IC-5.H.1:L10 Disposition of Separated UK Plutonium: A Review of Wasteform Options and Recent Progress
L.R. BLACKBURN, University of Sheffield, Sheffield, UK
In January 2025, following a substantial technical and economic analysis by the Nuclear Decommissioning Authority (NDA), UK Government announced a decision to implement an immobilisation strategy to manage its inventory of separated plutonium. This decision is the culmination of several iterations of strategic options analyses for a pathway to disposition that have considered, amongst other factors: technical feasibility and deliverability, socioeconomic benefit, market drivers and compatibility of feedstock with fuel manufacture. The NDA will now seek to improve its understanding of the physical, thermal, mechanical and corrosion properties of potential wasteform materials and identify a preferred immobilisation technology to implement disposal. At a high level, a hot isostatically pressed (HIP) zirconolite matrix, and a disposal mixed oxide (DMOX) wasteform are leading candidates as host phases for Pu disposition. This talk will outline recent developments in composition development, aqueous durability and radiation damage in zirconolite and DMOX wasteform simulants, and progress with regards to Pu validation.
IC-5.H.1:L11 Reduction of Spent Fuel Cladding Surface Oxide via Oxygen Dissolution into Metallic Zirconium using Spark Plasma Sintering
TOMOFUMI SAKURAGI, S. TANAKA, Radioactive Waste Management Funding and Research Center, Tokyo, Japan; H. HAYASAKI, S. MARUYAMA, Tokyo City University, Tokyo, Japan
Spent fuel claddings are planned to be disposed of deep underground. The metallic matrix of the cladding (Zr alloys) and the surface oxide film (zirconia; ZrO2) have different corrosion characteristics and radionuclide release behavior in geological environments. Zirconia is known to be a highly stable material. However, it would be promising to develop a uniform metallic matrix wasteform with a single component to avoid the complexity of post-closure safety assessments. Since the solid solution limit of oxygen in α-Zr is sufficiently high at around 28 at%, it may be possible to reduce the ZrO2 to Zr metal by incorporating the oxygen into the base metallic phase. In this study, Zr metal and ZrO2 powders were mixed to an oxygen concentration up to 25 at% and solidified using spark plasma sintering (SPS). XRD analysis revealed the extinction of the ZrO₂ peak. Cross-section analysis by EPMA showed a uniform distribution of oxygen in the 1700 °C sintered sample. This indicates that all the oxygen dissolved and diffused into the Zr metal phase, resulting in a Zr metal-only sintered compact. In conclusion, sintering the spent fuel cladding offers the prospect of reducing the oxide film to Zr metal, producing a metal-only ingot wasteform appropriated for final disposal.
IC-5.H.1:L12 Ceramic Wasteform and Process Development for Pu Disposition
A. FRISKNEY, M.C. STENNETT, L.R. BLACKBURN, University of Sheffield, Sheffield, UK
The UK holds around 140 tonnes of separated civil plutonium as a result of aqueous reprocessing of spent nuclear fuel. Following a UK government policy update (2025), civil plutonium is now intended for permanent disposition within a GDF. Pu will require immobilisation in a suitable wasteform; zirconolite-structured ceramics (CaZrTi2O7), processed via hot isostatic pressing, are a primary candidate on the basis of high aqueous durability and radiation stability. Uranium has been previously utilised as a Pu surrogate to evaluate wasteform compositions and inform future Pu trials. Here, we investigate a CuO process additive, previously demonstrated to reduce zirconolite formation temperature via liquid phase sintering. Facile formation of the target actinide host phase may reduce the requirement for high energy milling and carrier fluid in the process flowsheet, reducing the production of secondary waste streams. A suite of samples has been synthesised to assess how far the CuO additive can perform this function; U L3 edge XAS has been applied to understand the impact of HIP and process additives on the U speciation in the zirconolite phase. CuO addition is a novel approach contributing to the development and understanding of wasteform production at scale for safe Pu management.
Session IC-5.H.2 Challenging waste constituents, such as actinides, noble metals, and volatile species
IC-5.H.2:IL13 Thermochemistry of MXene used as Sorbent Materials for Off-gas Radioiodine Capture and Immobilization
K. BRINKMAN, Department of Materials Science and Engineering, Clemson University, Clemson, SC, USA
The capture and further immobilization of the major regulated volatile radionuclides (such as 129I and 85Kr) from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of significant research interest for the US Department of Energy (DOE) and other international colleagues. This work is focused on the development of MXenes as innovative sorbent materials for radioiodine capture and immobilization. MXenes have recently emerged as an intriguing family of graphene-like two-dimensional (2D) layer-structured transition metal carbides and/or nitrides with a high specific surface area, a tunable interlayer spacing, hydrophilicity, and abundant highly active surface sites. We used high-temperature oxidative oxide melt solution calorimetry, conducted at 800 °C in molten sodium molybdate (3Na2O·4MoO3), is used to determine the enthalpies of oxidation and formation from the elements at 25 °C, enabling a comprehensive thermodynamic assessment. These findings establish key correlations between crystal structure and energetic stability, providing valuable insights for the design of robust MXene-based materials capable of withstanding extreme environmental conditions. This work lays the thermodynamic foundation for the future development of MXenes in advanced energy and environmental technologies.
IC-5.H.2:IL14 How does Thermodynamics help evaluate Spent Fuel Options and Design Stable Nuclear Waste Forms
XIAOFENG GUO, Washington State University, Pullman, WA, USA
In the U.S. and worldwide, nuclear energy is advancing the decarbonization effort alongside renewables. A key challenge remains the management of spent nuclear fuel (SNF) and waste, which must be safely immobilized and disposed of. This requires a fundamental understanding of their structure, thermochemical stability, mechanical performance, and phase equilibria. Our group applies structural–thermodynamic approaches, particularly high-temperature calorimetry, to determine enthalpy quantities (heat capacity, phase transitions, reaction, mixing, and formation) to assess fuel and waste stability under relevant conditions. This talk summarizes how thermodynamic tools reveal the structure and stability of SNF and ceramic waste forms under storage and disposal conditions. Systems discussed include lanthanide-substituted UO₂ (simulated SNF or MOX) and geological-inspired ceramics such as zircon and brannerite. Beyond conventional UO₂, emerging fuels (e.g., uranium nitride, carbide, and silicide) and molten salts may generate new waste streams impacting fuel qualification and storage design. Our recent studies on UN and UC oxidation pathways under ambient and thermal conditions highlight how thermodynamic evaluations can guide waste management strategies for these next-generation fuels.
IC-5.H.2:IL15 AI-Accelerated Prediction of Glass Dissolution for Nuclear Waste Immobilization
N.M. ANOOP KRISHNAN, Department of Civil Engineering, Yardi School of Artificial Intelligence (Joint Appt.), Indian Institute of Technology Delhi, New Delhi, India
Predicting the long-term durability of nuclear waste glass remains challenging due to complex dependencies on composition, pH, temperature, and surface chemistry. In this talk, we will discuss how AI could be used to learn complex composition–property relationships of glasses in the context of nuclear waste immobilization. To this end, we developed an NLP-augmented machine learning model trained on ~700 literature datapoints that significantly outperforms traditional approaches by integrating compositional features with experimental conditions extracted via natural language processing. Our model demonstrates strong predictive accuracy and extrapolation capability to unseen compositions. By incorporating structural descriptors and applying SHAP analysis, we reveal underlying dissolution mechanisms and feature dependencies. This interpretable AI framework accelerates the discovery of durable glass compositions for sustainable nuclear waste management, demonstrating how domain-aware machine learning can bridge data-driven prediction with mechanistic understanding.
Session IC-5.H.3 Waste form modeling, performance testing, and advanced characterization techniques
IC-5.H.3:IL16 Interactions between Radionuclides and Organic Ligands with Cementitious Materials
YONGHEUM JO, Department of Nuclear Engineering, Hanyang University, Seoul, Republic of Korea
Cementitious materials are key waste forms and near-field engineered barriers in repositories for low- and intermediate-level waste (LILW). Infiltrating waters (groundwater or porewaters from groundwater–barrier reactions) contact and partially dissolve these materials, initiating radionuclide (RN) release. Thus, cementitious barriers both provide transport pathways (pore networks, fractures/cracks) and functions for RN isolation and retardation. Thus, Geochemical understanding of RN behavior in cementitious systems is essential: it supplies input data for performance assessment models and supports a credible safety case through validation against observations. LILW has lower radioactivity but larger volumes and more heterogeneous physicochemical properties. It also contains higher organic inventories from plant operations and decommissioning (e.g., chelates) and from degradation of polymeric wastes (e.g., cellulose, plastics). Such organics can act as ligands that increase RN solubility or suppress sorption, accelerating RN transport. They may also influence barrier degradation. This talk reviews RN–organic–cement interactions relevant to dose-significant LILW nuclides and shows how such insights strengthen near-field transport assessments and safety cases.
IC-5.H.3:IL17 Laser-Induced Breakdown Spectroscopy (LIBS) for the Characterization of Actinides Containing Waste
C. NOURRY, T. WISS, J. PRAVDOVA, European Commission, Joint Research Centre, Karlsruhe, Germany
Nuclear research institutions accumulate a wide range of legacy materials during their operational lifetimes, necessitating efficient and accurate characterization for safe management and disposal. Traditional analytical techniques such as mass spectrometry and scanning electron microscopy, while effective, are often time-intensive, laborious, and generate additional waste due to sample preparation and processing. At the Joint Research Centre Karlsruhe, researchers are exploring the application of Laser-Induced Breakdown Spectroscopy (LIBS) as a rapid analytical alternative for analysis of these materials. Preliminary studies have focused on non-irradiated nuclear fuels, including mixed oxide (MOX), uranium dioxide (UO₂), uranium nitride (UN) and uranium carbide (UC) demonstrating the technique’s reliability in characterizing radioactive, non-irradiated materials. The results highlight LIBS’s potential to streamline material sorting and reduce the need for conventional methods. Future work aims to adapt it for use in hot cell facilities, enabling the safe analysis of irradiated materials such as spent nuclear fuel and cladding. This advancement could significantly enhance the efficiency and safety of decommissioning and waste management processes in nuclear research and industry.
IC-5.H.3:IL18 Leaching Performance of Glass Waste Forms
J.W. AMOROSO, R.L. HAUSRATH, Savannah River National Laboratory, Aiken, SC, USA
Many classes of materials have been developed over the last 80 years to stabilize and treat nuclear waste for safe storage and disposal. Glass remains the predominant material of choice for radionuclide immobilization in applications where exposure and long-term stability are a priority. The suitability of glass is generally measured by accelerated leaching tests, the results of which are used to assess their performance. The data generated from accelerated leaching tests are used in model property predictions and to aid in compositional development. Glass concentrations high in Na2O and Al2O3 have a demonstrated propensity for devitrifying or precipitating crystals during cooling, which can adversely affect the performance. Correlation between physical and chemical changes with the thermal history of glass experienced during processing would facilitate performance predictions and allow for a wider range of durable compositions to be accessed. In this work, leaching tests conducted on a series of nuclear waste glasses to identify key correlations between chemical complexity, crystallization, and leaching performance. The results are used to develop a model for predicting crystallization in nuclear waste glass and ultimately towards a strategy for tailoring the glass waste form.
IC-5.H.3:IL19 Corrosion of Metallic Barriers in Nuclear Waste Disposal: Short Overview and Measurement & Monitoring Techniques
B. ZAJEC, Slovenian National Building and Civil Engineering Institute (ZAG), Ljubljana, Slovenia
The long-term integrity of metallic barriers is pivotal to the safety case of deep geological repositories for high-level nuclear waste. This talk will provide a concise overview of current canister designs featuring metallic walls, contrasting two engineering strategies: corrosion allowance materials (carbon steel, copper) and corrosion-resistant alloys (stainless steel, titanium-, and nickel-based alloys). Corrosion behavior in repository-relevant environments, as defined by national disposal concepts, will be summarized, highlighting uniform, localized (pitting, crevice), and stress corrosion cracking during the near-field evolution. Accurate determination of low corrosion rates and localized corrosion modes is essential for lifetime prediction. A selection of traditional and novel experimental techniques for assessing corrosion rates and characterizing degradation mechanisms will be presented, along with their advantages and limitations. Corrosion monitoring techniques will also be addressed, emphasizing in-situ field methods deployed in underground research laboratories to track canister performance under real repository conditions. These approaches will enable validation of predictive models and strengthen safety assessments over geological timescales.
Session IC-5.H.4 Materials issues in the design and operation of waste immobilization facilities
IC-5.H.4:IL20 Ceramics as Alternative Solutions for the Geological Disposal of High-level Radioactive Waste?
A. DEBELLE1, F. BUMBIELER1, G. GERMAN2,3, S. MIOT2, E. PERRET2, C. APHECEIXBORDE2, V. ALVAREZ1,4,5, C. BESNARD4, S. COUILLAUD4, C. ELLISALDE5, J.-M. HEINTZ5 , 1Andra, Scientific and Technical Division, Materials, Containers and HLW Disposal Cells Department, Châtenay-Malabry, France; 2Galtenco Solutions, Talence, France; 3IRT Saint Exupéry, Sites de Bordeaux et Toulouse, France; 4Ecole des Mines de St-Etienne, LGF, St-Etienne, France; 5IRCER, Université de Limoges / CNRS, Limoges, France
Andra is responsible for the Cigeo project, an Industrial Centre for Geological Disposal of radioactive waste, where high-level waste (HLW) will be disposed of at a depth of ~500 m in horizontal tunnels (cells) excavated in a claystone formation ensuring long-term confinement of radioactivity. For the shorter term, i.e. a few centuries, the confinement will also rely on the waste container enclosing the primary package. Currently, carbon steel is the reference material for both HLW containers and cell liners. However, under anoxic conditions, steel corrosion produces H₂(g). Gas generation is now addressed through complex design features, combining active and passive measures. To reduce this complexity while keeping safety requirements at a high level, alternative materials are being investigated—specifically ceramics, which do not release gas upon corrosion. In this framework, Andra and its partners are conducting R&D activities to identify, assess, and develop ceramic-based components. Two main options are under study: very thick monolithic alumina containers, and ceramic matrix composites for the cell liner. This presentation will provide an overview of the current status of these investigations.
IC-5.H.4:IL21 Radioactive and Toxic Waste Management: A Challenge or an Opportunity for New Hydraulic Binders?
X. BOURBON, Andra, Chatenay Malabry Cedex, France
Radioactive waste management is a worldwide preoccupation, leading to a significant number of high skilled research and surveys. The challenges regarding the downstream part of the nuclear fuel cycle cover subjects beyond the human scale and solutions need to be regularly reconsidered to take into account the evolution of our knowledge while ensuring long term safety. The main present solutions to confine ILW-LL (Intermediate Level-Long Lived waste), HLW (high Level Waste) as well as LILW-SL (Low Level-Short Lived Waste) are mainly based on the use of glass (for HLW), organic materials (e.g. epoxy resins) and in most of the cases cementitious materials (for low and intermediate level waste). Currently, the challenge is double: not only to new type of waste, still not handled yet or brand-new ones, have to be considered, but also the evolution of the availability of raw materials, especially cementitious ones, have to be taken into account. In this context, is this the opportunity to develop or characterize new binder? Do these new materials have to be simply based on the use of SCM’s (Supplementary Cementitious Materials) or considering a long-term management of radioactive waste, based on new and different matrix to define?