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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

The iodine-129 paradox in nuclear waste management strategies

Nuclear energy has an important role in the low-carbon energy transition, but the safety of spent nuclear fuel (SNF) management remains a public concern. Here, in this study, we investigate the interplay between waste management strategies and their environmental impacts with a particular focus on a highly mobile and persistent radionuclide, iodine-129 (I-129), which is the dominant risk contributor from SNF disposal and at existing groundwater contamination sites. The results show that the current recycling practice releases more than 90% of I-129 in SNF into the present-day biosphere using an isotropic dilution strategy, whereas the direct disposal of SNF in geological repositories is likely to delay and reduce the release by 8 orders of magnitude. In addition, our data synthesis of surface water concentrations near four nuclear facilities shows that the release-dilution strategy results in lower concentrations than regulatory standards, while insufficient waste isolation in the past has resulted in locally high concentrations within one site. Our analysis suggests that it is essential to consider effluents more explicitly as a part of the waste, that as society moves from dilution to isolation of waste, the potential risks of waste isolation to local regions should be carefully evaluated, and that excessive burdens of proof could hinder or discourage waste isolation. Comprehensive waste management strategies—considering not just volume but also mobility, isolation technologies and ultimate fates—are needed for persistent contaminants. This study offers valuable insights for optimizing the management of SNF and other persistent contaminants.

Wainwright, Haruko M. [Massachusetts Inst. of Tech↗
DOE OSTI2026

Peridynamic modeling of cementitious materials for nuclear waste management

Radioactive and hazardous waste generated from fuel processing plants, nuclear reactors, and hospitals, requires effective management strategies. Cementitious materials are widely applied for these needs, serving as structural materials, reactive barriers, or waste forms. In these applications cracking poses a significant risk to performance, driven by inconsistent shrinkage of the components and the varying strength and permeability of their interfaces. Traditional modeling approaches face challenges in representing the complex fracture behavior of cementitious materials due to the reliance on spatial derivatives and difficulties with mesh generation. Here, to overcome these limitations, we employ peridynamics, a novel continuum mechanics formulation that uses integrals to describe mechanical equilibrium, avoiding discontinuities associated with traditional methods. Through incorporation of a bi-linear softening model and quasistatics, an experimentally validated model for Portland cement concrete samples was created. Mechanical parameters, including compressive strength and elastic modulus were validated and variation due to aggregate packing was evaluated. Additionally, sensitivity analysis of the peridynamic parameters to the bulk material properties was established. The results lay the groundwork for evaluating the impact of unique conditions of cementitious waste forms that can be assessed to improve the reliability of waste management strategies.

Aggregates↗
DOE OSTI2025

Synthesis and characterization of rhenium chalcogenides for nuclear waste management

A new synthetic approach to binary rhenium (Re) chalcogenides was investigated to aid in the preparation of analogous technetium (Tc) compounds for Tc waste management efforts. The Boron-Chalcogen Mixture (BCM) method was utilized for the first time to prepare polycrystalline powders of ReS 2 , ReSe 2 , and Re 2 Te 5 using a perrhenate, NaReO4, as a rhenium source. Single crystals of ReS 2 were also synthesized using the combined BCM and molten flux methods by adding a K 2 CO 3 flux to the reagent mixture. Thermal analysis using TGA/DSC revealed negligible changes from ambient temperature up to ∼300 °C followed by the oxidation of the Re chalcogenides. High Temperature PXRD was used to investigate the crystallinity of ReS 2 at various temperatures revealing an increase in crystallinity up to 275 °C followed by a decrease in crystallinity due to oxidation. The research presented herein demonstrates a significant step towards technetium waste treatment efforts.

Boron chalcogen mixture (BCM) method↗
DOE OSTI2026

Transition Metals Separation with Commercial Neutral Extractants – A Review

The increasing use of extraction chromatography resins across fields such as hydrometallurgy, nuclear medicine, and environmental analysis has created a need for a deeper understanding of their interactions with transition metals. Despite extensive research on f-element separations, the behavior of transition metals in these systems remains relatively understudied. This review provides a comprehensive overview of the current state of knowledge on the extraction behavior of transition metals with neutral extractants, including TODGA, TEHDGA, TBP, and CMPO, and their corresponding resins, such as DGA, BDGA, UTEVA, TBP, and TRU. The review summarizes extraction data, extracted complex coordination environments, separation reaction stoichiometries, and associated thermodynamics, highlighting inconsistencies and knowledge gaps in the literature. The study emphasizes the need for further research using spectroscopy and computational methods to elucidate extraction mechanisms and to improve the efficiency and selectivity of transition metal separations. By identifying areas for future research and development, this review aims to stimulate advancements in the field and promote the development of innovative separation technologies. The implications of this research are far-reaching, with potential applications in nuclear waste management, nuclear forensics, metal recovery, and environmental remediation. Overall, this review provides a foundation for future studies on the extraction of transition metals using neutral extractants and resins.

Wall, Nathalie A.↗
DOE OSTI2026

Project Development of an Electrochemical Denitration and Caustic Generation System for HLW Pretreatment at Hanford - 26350

An engineering-scale electrochemical processing skid is proposed to perform the denitration of Hanford tank waste, which would help to mitigate a key process concern with the direct feed processing of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The reduction of nitrates and organic compounds in the waste feed will directly reduce hazardous NOx and ammonia gases generated during the vitrification process, which in turn will aid in addressing potential regulatory and safety challenges associated with processing large volumes of tank waste. This paper highlights the past legacy work, project layout, accomplishments from Phase 1 and research and development envisioned for Phase 2. An innovative electrochemical denitration and caustic generation (EDCGe) process was demonstrated for the pretreatment of tank waste at the Savannah River Site (SRS) in the early 2000s. The denitration electrolyzer, off-gas abatement system, and caustic generator electrolyzer are being developed with the intent that the denitration electrolyzer will convert nitrate and nitrite anions to nitrogen gas while also yielding other gaseous byproducts, which may include N2O, NH3, VOCs, and H2. The gaseous byproducts will be managed via a tandem off-gas catalyst-bed treatment system. The caustic generation electrolyzer will recycle NaOH from the feed to produce a clean caustic stream for use within the batching tanks at Hanford, aiding in the preparation of waste for WTP. The reduction in hazardous emissions and improved waste treatment processes provides a robust solution for nuclear waste management, contributing to environmental safety and regulatory compliance. The EDCGe technology is being adapted, modified, and updated for the preparation of the Direct Feed-High Level Waste (DF-HLW) flowsheet at Hanford. Phase 1 demonstrated a bench-scale proof-of-concept for reactions involving the denitration electrolyzer and gas phase abatement of ammonia. The electrochemical technology is drawing on the scientific outcomes that were reported in the legacy work. The results from Phase 1 demonstrated the viability of the EDCGe system in reducing the nitrogen species of simple non-radioactive waste simulants. Commercially available alloys used as electrode materials and membranes are being studied for the denitration and caustic generation electrolyzers. The continuation of this project holds promise for broader applications, such as energy-efficient ammonia production, and contributes significant advancements in nuclear waste management. Additional material discovery has been investigated into ceramic Na super ion conductive (NaSICON) materials and off-gas abatement catalyst discovery. NaSICON is of interest for selective transport of Na within the electrolyzers to make a clean caustic stream. Future integration and optimization efforts, informed by Phase 1 results and ongoing research, will continue to drive advancements in nuclear waste management technology. The technology developed for the EDCGe treatment of tank waste will also have broader potential to inform other fields, such as energy-efficient ammonia production, as well as ammonia abatement catalysis through the lessons learned in electrochemical nitrate reduction. The applications and benefits of this research extend beyond Hanford and the Savannah River Site, supported by a collaborative team of scientists and engineers from national labs, academia, and industry, ensuring a comprehensive approach to solving complex waste treatment challenges. The team is leveraging advanced electrochemical technologies, machine learning, novel catalysts tailored for gaseous nitrogen species, and cutting-edge reactor systems to enhance the process efficiency and effectiveness of the denitration process.

Rodene, Dylan [Savannah River National Laboratory ↗
DOE OSTI2025

Guiding Principles for Geochemical/Thermodynamic Model Development and Validation in Nuclear Waste Disposal: A Close Examination of Recent Thermodynamic Models for H + —Nd 3+ —NO 3 - (—Oxalate) Systems

Development of a defensible source-term model (STM), usually a thermodynamical model for radionuclide solubility calculations, is critical to a performance assessment (PA) of a geologic repository for nuclear waste disposal. Such a model is generally subjected to rigorous regulatory scrutiny. In this article, we highlight key guiding principles for STM model development and validation in nuclear waste management. We illustrate these principles by closely examining three recently developed thermodynamic models with the Pitzer formulism for aqueous H + —Nd 3+ —NO 3 - (—oxalate) systems in a reverse alphabetical order of the authors: the XW model developed by Xiong and Wang, the OWC model developed by Oakes et al., and the GLC model developed by Guignot et al., among which the XW model deals with trace activity coefficients for Nd(III), while the OWC and GLC models are for concentrated Nd(NO 3 ) 3 electrolyte solutions. The principles highlighted include the following: (1) Principle 1. Validation against independent experimental data: A model should be validated against experimental data or field observations that have not been used in the original model parameterization. We tested the XW model against multiple independent experimental data sets including electromotive force (EMF), solubility, water vapor, and water activity measurements. The results show that the XW model is accurate and valid for its intended use for predicting trace activity coefficients and therefore Nd solubility in repository environments. (2) Principle 2. Testing for relevant and sensitive variables: Solution pH is such a variable for an STM and easily acquirable. All three models are checked for their ability to predict pH conditions in Nd(NO 3 ) 3 electrolyte solutions. The OWC model fails to provide a reasonable estimate for solution pH conditions, thus casting serious doubt on its validity for a source-term calculation. In contrast, both the XW and GLC models predict close-to-neutral pH values, in agreement with experimental measurements. (3) Principle 3. Honoring physical constraints: Upon close examination, it is found that the Nd(III)-NO 3 association schema in the OWC model suffers from two shortcomings. Firstly, its second stepwise stability constant for Nd(NO 3 ) 2+ (log K 2 ) is much higher than the first stepwise stability constant for NdNO 3 2+ (log K 1 ), thus violating the general rule of (log K 2 –log K 1 ) < 0, or $\frac{K1}{K2}$>1. Secondly, the OWC model predicts abnormally high activity coefficients for Nd(NO 3 ) 2 + (up to ~900) as the concentration increases. (4) Principle 4. Minimizing degrees of freedom for model fitting: The OWC model with nine fitted parameters is compared with the GLC model with five fitted parameters, as both models apply to the concentrated region for Nd(NO 3 ) 3 electrolyte solutions. The latter appears superior to the former because the latter can fit osmotic coefficient data equally well with fewer model parameters. The work presented here thus illustrates the salient points of geochemical model development, selection, and validation in nuclear waste management.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI1996

Nuclear Safety [Vol. 37, No. 3, July-September 1996]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS 181 Attitudes and Practices Regarding Disposal of Liquid Nuclear Waste at Clinton Laboratories in the Very Early Years: A Historical Analysis, S. H. Stow; 202 Off-Site Nuclear Emergency Management in Germany Under the Auspices of the Federal Structure, A. Bayer, S. Bittner, and H. Korn; DESIGN FEATURES: 211 Twenty-Fourth DOE/NRC Nuclear Air-Cleaning and Treatment Conference, R. R. Bellamy, J. J. Hayes, R. D. Porco, and M. W. First; OPERATING EXPERIENCES: 222 Explosion in the Tomsk-7 Reprocessing Plant on April 6, 1993, M. L. Hyder, W. G. Lussie, and F. E. Witmer; 234 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 237 Assessment of Spent Fuel Cooling, J. G. Ibarra, W. R. Jones, G. F. Lanik, H. L. Ornstein, and S. V. Pullani; RECENT DEVELOPMENTS: 256 Reports, Standards, and Safety Guides, D. S. Queener; 260 Proposed Rule Changes as of June 30, 1996; ANNOUNCEMENTS: 210 Tennessee Industries Week; 221 21st International Symposium on The Scientific Basis for Nuclear Waste Management; 259 IAEA Programme of Scientific Meetings for 1997; 270 PSAM 4 (International Conference on Probabilistic Safety Assessment and Management); 265 The Authors; 268 Errata.

05 NUCLEAR FUELS↗
NASA NTRS1977

An analysis of the technical status of high level radioactive waste and spent fuel management systems

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

English, T.↗
DOE OSTI2025

Reevaluation of Radiation-Protection Standards for Workers and the Public Based on Current Scientific Evidence

President Trump’s recent executive orders to “Usher in a Nuclear Renaissance,” coupled with the global pledge to triple nuclear energy capacity by 2050, underscore nuclear energy’s importance to national security and economic prosperity. This renewed interest has prompted efforts to spur nuclear-energy deployment, including assessing factors impeding it. One issue that has previously been identified as adding to the cost of nuclear energy is excessively conservative requirements, including those related to radiation protection. This technical review, therefore, examines current radiation-protection standards that were established decades ago when more limited data were available and nuclear-energy expansion was not a national priority. The review focuses on scientific evidence regarding the health effects of ionizing radiation at annual doses of 10,000 mrem or less. The review evaluates epidemiological studies, radiobiological research, and positions of relevant professional organizations to assess whether such doses result in discernable or observable increases in negative health outcomes. The review also surveys the literature related to economic and practical implications of current radiation-protection standards and practices. Based on this assessment, we propose maintaining an annual occupational whole-body dose limit of 5,000 mrem/yr and eliminating all “as low as reasonably achievable” requirements and limits below this threshold. This change could potentially reduce radiation-protection costs by millions of dollars annually for each reactor, as well as decrease the overall costs and correct misconceptions about the risks associated with all nuclear technologies. The evidence further supports future consideration of a 10,000 mrem/yr limit that would maintain appropriate safety margins while further reducing protection costs. Similarly, given the data and that the average annual radiation dose per person in the U.S. is 620 mrem, we believe the public dose limits of 100 mrem/yr are unnecessarily restrictive; increasing to 500 mrem/yr would maintain substantial safety margins—a factor of 10 below the occupational limit—while reducing regulatory burdens and associated bureaucracy. While we acknowledge ongoing scientific debate and encourage continued research on the health effects of ionizing radiation, our review indicates current frameworks are overly conservative. These overly stringent limits not only impose unnecessary economic burdens without corresponding health benefits but also divert safety focus and resources from more important considerations. Although this study was motivated by nuclear-power considerations, reforms to radiation-protection requirements have significant positive implications for other areas, such as nuclear medical applications, environmental remediation, nuclear-waste management and disposal, and industrial applications of nuclear technologies.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2026

Hydrogen Recombiner Catalyst Evaluations for Waste Storage

Radiolysis of water in nuclear waste storage generates hydrogen gas that can accumulate within sludge style waste and be rapidly released during agitation events, creating a significant flammability hazard. Engineering controls are therefore required to limit hydrogen concentrations during both quiescent storage and transient disturbances. Catalytic recombination of hydrogen in waste storage offgas is a proven mitigation strategy, maintaining hydrogen levels below flammability limits and managing sudden concentration spikes. Conventional recombiners rely on platinum and/or palladium catalysts, with development efforts focused on extending service life, increasing active surface area, and ensuring safe deployment in radioactive environments. Savannah River National Laboratory (SRNL) is evaluating a newly developed hydrogen recombiner catalyst from Canadian Nuclear Laboratories as a cost-effective and durable alternative for nuclear waste applications. Testing was conducted in SRNL’s Shielded Cells facility, which enables reduced-scale experimental modeling under radiation fields and near-use-case conditions relevant to radioactive waste storage. Catalyst performance was evaluated using a custom offgas characterization system designed for near-zero flow conditions. The experimental apparatus consisted of a gas-tight 2.7 L PTFE vessel equipped with temperature monitoring, gas flow controls, and a variable-speed mixer to simulate sludge agitation. Offgas composition was monitored using a dedicated gas chromatograph with argon carrier gas and a krypton internal standard. Measurements were obtained for an empty vessel, the vessel containing a well characterized radioactive tank waste sample, and the same configuration with the candidate catalyst installed. Results demonstrate that the new catalyst effectively reduced hydrogen concentrations in the offgas within the constraints of the experimental design. In addition to confirming catalytic activity, the testing provided valuable insights into experimental optimization and considerations for future performance evaluations. These findings support the potential scalability of the technology and highlight its applicability to broader nuclear waste management operations, offering improved safety and reduced operational costs through enhanced catalyst durability and lower replacement frequency.

Tener, Zachary P. [Savannah River National Laborat↗
DOE OSTI2026

Localizing tetrahedral aluminum in nitrate-bearing gibbsite to constrain defect-impurity coupling

The enhanced radiolytic stability of gibbsite (α-Al(OH) 3 ) containing trace nitrate (NO 3 − ) is a phenomenon in nuclear waste management, but its structural origins remain unresolved. Motivated by the detection of minority tetrahedral aluminum (T d ) defects in synthetic gibbsite, we hypothesized that these sites may participate in NO 3 − retention or mediate H 2 suppression. To evaluate this, we combined orthogonal techniques comprised of spatially selective solid-state 27 Al MAS NMR, comparative spectroscopy, and density functional theory (DFT) modeling. Paramagnetic editing and dynamic nuclear polarization (DNP) MAS NMR confirm that T d defects are confined to the particle interior. DFT calculations reveal no energetic stabilization of NO 3 − near T d sites. Comparative NMR analysis shows that T d is also present in chloride-bearing gibbsite, which exhibits high radiolytic hydrogen yields. These three independent disqualifications rule out T d as a structural contributor to nitrate-mediated suppression and narrow the scope of defect-driven explanations. The findings redirect mechanistic attention away from coordination defects and toward redox-active impurity pathways, providing a refined foundation for understanding radiation tolerance in Al(OH) 3 .

Graham, Trent R. [Pacific Northwest National Labor↗
DOE OSTI2025

Geobacter sulfurreducens Immobilized Boron-Doped Diamond Electrodes for Uranium(VI) in Water Electrochemical Bioremediation

The proliferation of nuclear science and technology has resulted in an increase in nuclear waste containing uranium, posing significant risks to both human and environmental health. This study proposes the use of Geobacter sulfurreducens (G. sulfurreducens) modified boron-doped diamond electrodes to facilitate the reduction and removal of uranium(VI) from aqueous media. The bioremediation process involves electrochemically immobilizing the bacteria on a boron-doped diamond electrode (BDD). The immobilization process requires applying reduction potentials ranging from −0.40 to −0.70 V (vs Ag/AgCl (3 M NaCl)), with −0.60 V identified as the optimal potential for effective bacterial modification. The uranium source is provided by a 2.0 mM uranyl acetate solution in G. sulfurreducens growth medium. Scanning electron microscopy (SEM) reveals a highly uniform layer of uranium on the electrode surface. Energy-dispersive X-ray fluorescence spectroscopy (EDS) and cyclic voltammetry (CV) studies confirm the presence of uranium in the system. Raman spectroscopy and X-ray photoelectron spectroscopy (XPS) successfully elucidate the reduction process of U(VI) to predominantly U(IV) using a bacteria-electrode coupled system. Additionally, a comparison is made with the electrochemical removal of uranyl ions using the electrodeposition method on unmodified BDD. Results demonstrate the presence of three uranium oxide species (UO 2 , UO 3 , and U 3 O 8 ) on the BDD electrode after experimentation, in contrast to the G. sulfurreducens/BDD assembly, which achieves the predominant reduction of UO 2 2+ to UO 2 with a small quantity of UO 3 as the final species. This study highlights the efficient electrochemical removal of uranyl ions from aqueous media at the G. sulfurreducens/BDD interface through chronoamperometry, presenting a promising approach for remediating sites contaminated with radioactive materials. The findings contribute to the exploration of sustainable alternatives for managing nuclear waste, emphasizing the potential of this electrochemical bioremediation strategy.

Bacteria↗
DOE OSTI2025

Toward integrating high-fidelity CFD approaches in the thermal-hydraulic analysis of turbulent dry cask systems

Nuclear power plants have been supplying resilient and reliable electricity for decades, contributing to energy independence of the U.S.. However, nuclear waste management remains one of the most significant challenges in the industry. The safety of dry cask storage systems relies heavily on their thermal-hydraulic performance. Computational Fluid Dynamics (CFD) simulations are often used to demonstrate this performance and ensure that the system design meets safety standards. This study presents reduced numerical models for various types of dry cask systems. These numerical models can produce efficient and fast results based on the employed modeling strategies. Additionally, the study uses a novel approach to high-fidelity simulations to evaluate modeling assumptions in dry cask modeling. Large Eddy Simulations (LES) are used for this purpose, particularly in regions where fluid velocity is relatively high and the turbulence characteristics become important. Furthermore, the results of these high-fidelity simulations will enhance the interpretation of outcomes produced from a lower-fidelity CFD model.

CFD↗
DOE OSTI2025

Discrete fracture network model benchmarks developed and applied in a DECOVALEX-2023 repository performance assessment study

This study presents newly developed benchmarks for modeling flow and transport within discrete fracture networks (DFNs) and useful methods for analyzing the results. The new benchmarks are designed to test modeling approaches for use in probabilistic performance assessment models of deep geologic repositories in fractured rock. The benchmarks simulate flow and transport through a 1 km 3 block of fractured rock. The first simulates migration of a short pulse of tracer through a simple network of four intersecting fractures. The second adds 1089 stochastically generated fractures. The third changes the pulse to a continuous point source. Evaluation of model performance relies on moment analysis and comparison of the results of different models. The expected nondimensional first moment of the conservative tracer for each benchmark is 1. The benchmarks were simulated by teams from Canada, Czechia, Germany, Korea, Sweden, Taiwan, and the United States as part of a DECOVALEX-2023 study (decovalex.org). The teams used various approaches, including explicit DFN modeling, DFN upscaling to an equivalent continuous porous medium (ECPM), and a combination of both methods. Transport mechanisms are modeled using either the advection-dispersion equation or particle tracking. Results demonstrate strong agreement among the models in breakthrough behavior up to the 75th percentile. Significant deviations in first moments and well-clustered outputs led to the identification of inaccuracies in several models. Such findings exemplify the benefit of exercising these benchmarks and using the presented methods to test DFN flow and transport models.

Benchmark↗
DOE OSTI2026

Overview of Wendelstein 7-X high-performance operation

The Wendelstein 7-X (W7-X) stellarator has completed two consecutive experimental campaigns OP 2.2 (Sep.-Dec. 2024) and OP 2.3 (Feb.-May 2025) under a new operational strategy enabling more than one year of uninterrupted device availability. This approach, supported by exceptionally high subsystem reliability, allowed sustained high-efficiency plasma operations with up to 80–100 discharges per day across a broad range of magnetic configurations. Several key technical upgrades-most notably the first operation of a 1.5 MW class steady-state gyrotron, a new steady-state pellet injector, and advanced real-time feedback control systems significantly enhanced heating, fueling, and plasma control capabilities. Together, these improvements enabled major advances in long-pulse performance, high-β operation, and confinement optimization. Long-pulse discharges achieved 1.8 GJ of injected energy under fully detached divertor conditions, while reduced-field scenarios facilitated record volume-averaged β values approaching 3%. High-performance plasmas with centrally peaked density profiles, created via neutral beam injection (NBI) or sustained pellet fueling, demonstrated strongly reduced turbulent transport and stellarator-record fusion triple products. Complementary studies of power exhaust and divertor heat loads revealed the role of scrape-off-layer drift physics in shaping strike-line patterns under attached conditions. Together, the results from OP 2.2 and OP 2.3 significantly expand the operational space of W7-X and strengthen its role as a leading platform for steady-state stellarator research and reactor-relevant plasma scenarios.

Wendelstein 7-X↗
DOE OSTI2025

Perspective Chapter: Safe Disposal and Storage of Nuclear Waste

The use of nuclear energy inevitably generates nuclear waste as the byproduct of fission reactions. Depending on the initial composition of the fuel that goes into the reactor and the subsequent burn-up level, the chemistry of the resulting nuclear waste can vary substantially. This waste typically exhibits a broad spectrum of radioactivity and half-lives, making effective management one of the most critical challenges for global nuclear energy. This chapter provides a comprehensive overview of the origin and classification of nuclear waste and various strategies for its safe immobilization and disposal. The short- and long-term storage of waste with varying radioactivity is addressed. The significant technical and political complexities involving primarily long-term disposal are also discussed. To ensure the safe and permanent disposal of hazardous waste with extremely long half-lives, future efforts should focus on both technical innovation and public engagement.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Recent Advances in Cladding Material Extraction from Fuels in Nuclear Fuel Cycles

An improved recycling and recovery process for the cladding material from spent nuclear fuels is very important toward confirming nuclear energy to support ongoing sustainable development of nuclear management by reducing waste and conserving resources. Nuclear spent fuel cladding materials such as zirconium alloys have economic values and can be recovered, and their recovery eliminates problems in waste disposal and conserves valuable resources. Over 110 published reports and journal articles are reviewed and summarized herein, with a main focus on documenting recovery techniques used to recover cladding materials from spent nuclear fuel and recent developments. Several recovery techniques which are used at present times, such as mechanical separation, chemical dissolution, and hydrometallurgical processes have been covered with examples and discussions. Difficulties within the recovery process are also discussed, and most probable areas for future research in improving efficiency and sustainability of recovering cladding material are identified and discussed at the end. Here, this review could be an important document to the field of spent nuclear fuel reprocessing, recovering valuables and thereby offering guidance on how to effectively manage, safely handle, and reduce nuclear waste. In addition to reducing the volume and radiotoxicity of high-level waste, this review also highlights the potential economic benefit of recovering zirconium from spent fuel cladding by relating typical zirconium metal prices to the mass of cladding per tonne of spent fuel, illustrating that the recoverable material value is non-negligible compared with back-end fuel-cycle costs.

Mondal, Kunal [Oak Ridge National Laboratory (ORNL↗