TIR-Raman Sensor for Characterizing Nuclear Waste Slurries
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Hanford Tank Waste Technology Development Integration and Innovation Synergy
The Chief Technology Office (CTO) within Hanford Tank Waste Operations & Closure (H2C) identifies technology development needs and develops solutions for tank waste treatment. One key element of the technology development (TD) program is the annual Technology Development Integrated Coordination (TDIC) meeting series. The main goal of these meetings is to inform the National Laboratories and other entities of the needs identified by the end users of the technology. In turn, the meetings provide an opportunity for National Laboratories and others to provide insight into potential solutions. This dynamic facilitates dialogues that frequently identify not only new, innovative approaches, but also enhancements to ongoing technology development projects.
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Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal
Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Accelerating Nuclear-Integrated Data Centers in the USA: SWOT Analysis, Power-Thermal Management Strategies, and Industrial-Scale Demonstration and Potential Deployment
Driven by the growth in digital services, cloud computing, AI, and manufacturing, data centers face rising energy demands that challenge traditional power sources and cooling efficiency. This study explores using nuclear power to meet these demands, focusing on accelerated reactor technology deployment and highlighting needs such as N+1/N+2 power supplies and integrated power-thermal management. A SWOT analysis addresses grid connectivity, reactors, and site selection, particularly DOE sites. Reactor technology demonstration and deployment could be accelerated by leveraging test facilities such as MARVEL, MAGNET, TED, FAS, DOME, LOTUS, ATR, Energy System Proving Grounds, and upcoming Energy Launch Pads, along with modeling and simulation tools such as RELAP5, MOOSE, VERA, RAVEN, and FORCE. The potential power and thermal management options, including various cooling technologies, waste-heat utilization, and an industrial-scale demonstration plan, aim to accelerate the integration of nuclear power and data centers in the USA, while emphasizing community and stakeholder engagement and synergistic efforts.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Nanohybrid of Silver‐MXene: A Promising Sorbent for Iodine Gas Capture from Nuclear Waste
The increasing reliance on nuclear energy as a significant low-carbon power source necessitates effective solutions for managing radioactive emissions. This study introduces a novel application of MXene nanohybrids, specifically silver-MXene (Ag-Ti 3 C 2 T x ), as an effective sorbent for radioiodine off-gas capture at an operating temperature of 150 °C. Through comprehensive material characterization, including X-ray diffraction, scanning and transmission electron microscopies, energy-dispersive X-ray spectroscopy, Raman spectroscopy, thermogravimetric analysis, inductively coupled plasma optical emission spectroscopy, and gas sorption analyses, the successful loading of Ag nanoparticles onto Ti 3 C 2 T x is confirmed and the subsequent formation of AgI upon iodine capture. The results demonstrate that Ag-Ti 3 C 2 T x exhibits superior iodine uptake compared to traditional silver-based sorbents such as silver mordenite zeolite (AgZ) and silver-functionalized silica aerogel (AgAero). The Ag-Ti 3 C 2 T x achieves an iodine loading of 946 mg g −1 , significantly outperforming AgZ (131 mg g −1 ). These findings highlight the potential of Ag-Ti 3 C 2 T x as a highly efficient, thermally stable sorbent for radioiodine capture, and potentially addressing key limitations of existing materials.
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Segmented Gamma Scanner for Radioactive Waste Assessment: A User Guide
Radioactive waste assessment is important for ensuring nuclear material security at various types of facilities, such as enrichment, fuel fabrication, and reprocessing plants. The waste generated at such nuclear facilities is stored in standard containers and is required to be characterized for material-accounting purposes. The segmented gamma scanner system is a popular, nondestructive analysis measurement system used for characterizing nuclear material, including radioactive waste. This document provides guidance on how to achieve effective performance from a segmented gamma scanner system for accurately quantifying fission products, activation products, and transuranic wastes.
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Full Scale HEPA Filter Encapsulation in Ultra-High-Performance Grout - Proof of Concept
The Hanford Tank Waste Treatment and Immobilization Plant (WTP) currently being constructed to treat radioactive waste, includes vitrification facilities for both the high-level waste (HLW) and low activity waste (LAW) fractions. Operation of the WTP will produce contaminated high-efficiency particulate air filters (HEPA), as part of the solid secondary waste (SSW) stream. The HEPA filters receive off-gas from the vessel vent header and primary off-gas treatment system in LAW Facility and remove particulate contaminants including 99 Tc and 129 I salts. The current disposal method for HEPA filters is encapsulation in metal containers using cementitious material (CM) and disposal in the Integrated Disposal Facility (IDF). Results from the 2017 IDF Performance Assessment (PA) WRPS (2018) demonstrated that while compliance is maintained for the 1000 year compliance period mandated by DOE O 435.1 and its accompanying manual, release of constituents from the HEPA filters result in exceedance of the performance objective imposed as the groundwater regulatory limit at later times. For example, at about 1500 years post-closure, solid secondary waste (SSW), including HEPA filters is predicted to become a dominant contributor to 99 Tc release and over the 10,000-year sensitivity analysis period, SSW is the dominant contributor of 129 I release to the groundwater. The estimated release could potentially be reduced if the HEPA filters are not compacted and waste containers could be distributed throughout a large space, thereby diluting the contaminant release. Additionally, a better cementitious material could be used to encapsulate the HEPA filters. One alternative method for disposal of contaminated HEPA filters is encapsulation of the filters in ultra high-performance grout (UHPG). UHPG is a variation of ultra-high-performance concrete (UHPC) is commonly used in the prestressed concrete industry for large structural members. Recent studies of UHPG show it has excellent properties for containing radionuclides such as 99 Tc and 129 I Nichols and Kaplan (2021). This report presents the results of the first attempt to encapsulate a clean, full-size HEPA filter in UHPG and evaluate the effectiveness of the immobilization process and final waste form. A full-scale proof-of-concept simulated waste form was prepared by encapsulating a HEPA filter in a 110-gallon stainless steel (SS) drum using UHPG. A change from Type I/II PC to Type 1L PLC was made after American Rock Products informed the team that they would no longer be using Type I/II by the end of 2024 and the northwest was phasing out Type I/II PC overall. Type I/II PC used in previous studies of UHPG for encapsulation (Nichols and Kaplan 2021). After the UHPG was cured both the scaled mockup and the full-scale simulated waste forms were sectioned for visual examination. UHPG completely encapsulated the filters and bonded to the external surfaces of materials comprising the filters. No cracks were observed in the sectioned waste forms.
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Viscosity Measurements of the HLW APPS Glasses
The U.S. Department of Energy is responsible for building the Hanford Waste Treatment and Immobilization Plant (WTP) at the Hanford site in Washington to remediate 56 million gallons of radioactive waste historically stored in 177 underground tanks. The Office of River Protection has requested that the Savannah River National Laboratory (SRNL) contribute in areas of recognized capabilities and expertise for glass waste form development to support successful startup of the WTP. SRNL support of this work is defined in the Task Technical and Quality Assurance Plan.1 This report provides results from viscosity measurements on a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory (PNNL). The glasses were designated the high-level waste – Aspen Process Performance Simulation (HLW-APPS) study glasses. The data provided in this report is to be used in the development, validation, and implementation of enhanced property/composition models for nuclear waste glasses.
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Development of a Multi-Robot System for Autonomous Inspection of Nuclear Waste Tank Pits
This paper introduces the overall design plan, development timeline, and preliminary progress of the Autonomous Pit Exploration System project. This project aims to develop an advanced multi-robot system for the efficient inspection of nuclear waste-storage tank pits. The project is structured into three phases: Phase 1 involves data collection and interface definition in collaboration with Hanford Site experts and university partners, focusing on tank riser geometry and hardware solutions. Phase 2 includes the selection of sensors and robot components, detailed mechanical design, and prototyping. Phase 3 integrates all components into a cohesive system managed by a master control package which also incorporates digital twin and surrogate models, and culminates in comprehensive testing and validation at a simulated tank pit at the Idaho National Laboratory. Additionally, the system’s communication design ensures coordinated operation through shared data, power, and control signals. For transportation and deployment, an electric vehicle (EV) is chosen to support the system for a full 10 h shift with better regulatory compliance for field deployment. A telescopic arm design is selected for its simple configuration and superior reach capability and controllability. Preliminary testing utilizes an educational robot to demonstrate the feasibility of splitting computational tasks between edge and cloud computers. Successful simultaneous localization and mapping (SLAM) tasks validate our distributed computing approach. More design considerations are also discussed, including radiation hardness assurance, SLAM performance, software transferability, and digital twinning strategies.
Nuclear waste management↗
Unveiling the Role of Critical Impurities in Spent LiFePO 4 Cathodes for Scalable Direct Regeneration
Direct regeneration offers a promising alternative to recycling End-of-Life (EoL) batteries by restoring metal elements and preserving the material structure, yet scaling these technologies to handle practical cathode black mass (CBM) with impurities remains challenging. Here, this study investigates the evolution of impurities, including aluminum (Al), polyvinylidene difluoride (PVDF) binder, and residual carbon (C), during direct recycling of spent LiFePO 4 (LFP) cathodes and their impact on electrochemical performance. Using various ex situ and in situ analyses, it is shown that the formation of lithium fluoride (LiF) during the traditional direct recycling process hinders lithium diffusion and deteriorates the reversible capacity. To address this major challenge, the combination of pH-controlled hydrothermal purification and the two-step sintering process is proposed effectively to regenerate spent LFP cathodes, eliminating the negative effect of Al and fluorine (F) impurities while mitigating any potential impacts of carbon residuals. The regenerated LFP from spent CBM achieves superior performance, retaining 152.5 mAh g −1 at 0.1 C and 133 mAh g −1 at 1 C with 98.7% capacity retention after 200 cycles. This approach is further validated using three distinct waste feedstocks from battery modules, enhancing impurity management and scalability in direct recycling. These findings present a sustainable and economically viable solution for large-scale LFP regeneration.
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Performance Assessment for the Disposal of Low-Level Waste in the 200 West Area Burial Grounds (Annual Summary Report - FY 23)
This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 West Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. All the trenches in these burial grounds except two have remained inactive since 2005. The only active trenches that continue to receive low-level and mixed low-level waste are Trenches 31 and 34 that are located in the southern portion of the 218-W-5 Burial Ground. During this reporting period (fiscal year 2023, from October 1, 2022, through September 30, 2023), waste was disposed in these two trenches.
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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).
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Simulant Development of Potential 200 West Area Waste Feeds
Preliminary planning for retrieval, qualification, and pretreatment of waste in Hanford’s 200 West Area (200W) has begun as part of the West Area Risk Management project. Experimental studies to technically mature pretreatment process operations will likely be needed because of the uniqueness of 200W waste. Pacific Northwest National Laboratory formulated five simulants to represent 200W-qualified feed based on the preliminary flowsheet provided by Washington River Protection Solutions, LLC. The simulant recipes were devised using applicable historical information as a reference point to support the use of the flowsheet waste vectors, which were combined into five distinct groups. These five groups formed the basis for the liquid composition targets that were adapted into recipes using charged-balanced salt species. The liquid phase recipes were batched in 1-L quantities and analyzed at Pacific Northwest National Laboratory. Once confirmed to be stable, the liquid solutions were tested for compatibility with candidate solid components. Specific solid components were recommended based on cross-examining the proposed solid phases in the flowsheet with relevant data from the literature. Mixtures of solid components were added to aliquots of the liquid batches and sub-sampled to measure particle size distribution. The measured distribution was compared to independently created benchmark distributions appropriate for each simulant. This process was iterated until a solid phase composition that resulted in a representative particle size distribution was found. After the final compositions were confirmed, a suite of chemical and physical characterization data was collected. This report describes the simulant basis, formulation methodology, laboratory measurements, and data collected for the recipes recommended to represent 200W waste feeds.
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Nevada National Security Site Waste Disposal Operations - Volume Report FY 2024 - Quarter Two
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Nevada National Security Site Waste Disposal Operations - Volume Report FY 2024 - Quarter Three
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International Collaboration Activities in Geologic Disposal Research (FY24 Progress Report)
This report describes the FY24 status of international collaboration regarding geologic disposal research in the Office of Spent Fuel and Waste Science and Technology (SFWST) in the Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of SFWST is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal.
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International Collaboration Activities in Geologic Disposal R&D: Spent Fuel and High-Level Waste Disposition (FY25 Progress Report)
This report describes the FY25 status of international collaboration on geologic disposal research and development (R&D) in the Office of Spent Fuel and High-Level Waste Disposition (SFHLWD) in the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of this office is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal. R&D in geologic disposal is administered in SFHLWD’s Disposal R&D Campaign.
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Microwave-assisted catalytic gasification of mixed plastics and corn stover for low tar, hydrogen-rich syngas production
The challenge for efficient management of post-consumer plastic and biomass waste has grown over the past few decades due to their dramatic increases. In comparison to conventional gasification, microwave-assisted co-gasification of plastics and corn stover offers many benefits, including increased H 2 yield and gas components compared to unfavorable char/tar. Nonetheless, for future commercialization of the process and ease of product separation, further reduction of the undesirable tar is necessary, which can be achieved over the catalytic route. Here, in this work, we studied the catalytic effect of magnetite for microwave-assisted co-gasification of corn stover and plastic to make syngas with higher H 2 and lower tar selectivity over non-catalytic conditions. A 1:1:1 ratio of plastic-corn stover-magnetite was used to evaluate the reaction parameters such as temperature, space velocity, heating media, and catalytic cycles under gasification conditions. In comparison with the microwave non-catalytic route, a 100% increase in the total H 2 yield with 76% higher H 2 production efficiency (mmol/kWh) was achieved in the presence of the magnetite catalyst, while reducing the overall tar formation from 9% to 2%. When magnetite was reduced in situ during the reaction, it coupled with microwave and delivered oxygen radicals that cracked down plastic and corn stover intermediates generated from the synergistic effect under microwave heating. Soon after the oxygen transfer process initiated, magnetite reached its final oxidation state consisting of microwave-active Fe and Fe 3 C phases that continued coupling with microwaves along with the generated graphitic carbon to maintain the heat necessary to further reduce the generated tar and make additional gaseous products, as confirmed by XRD, Raman, and TGA analyses.
08 HYDROGEN↗