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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,045 records · Page 58

EM-Enhanced HyPOR Loop for Fast Fusion Cycles

Vacuum pumps are the heart of a fusion energy facility – fusion power cannot be generated without them. Nevertheless, the vacuum technology needed to operate a viable fuel cycle for a compact fusion power plant does not exist. Commercial vacuum technology offers the best solution to this challenge, but a pump oil recycling and detritiation system is necessary. Conventional oil detritiation processes have only been developed to deal with legacy waste and are too slow and destructive to the oil. Further, post hoc detritiation strategies are intrinsically inefficient. Our approach is to rethink the challenge holistically by designing the pumping fluid and tritium extraction system in an integrated manner to achieve an innovative solution. By selecting an oil for the specific task of pumping tritium and then designing our catalytic system to selectively target the tritium-bearing functional groups, an effective detritiation system (hydrocarbon pump oil recycling, HyPOR, loop) for fusion power plants can be obtained. The project demonstrated a HyPOR loop process that can selectively remove heavier hydrogen isotopes from pump oil, reaching the target of 99.5 % removal, with an uptake of less than 0.01% of tritium throughput, while also purifying the oil of radiation-induced damage. The recycled oil retained its pumping characteristics over 7 recycles and gamma irradiation over 30 MGy. By meeting these targets, the project enable a reduction in pump operational costs from $\$$14.5M/year to $\$$103k/year (>140× reduction), reducing pump electric power consumption from 2.8 MW to 0.25 MW (>10× reduction), and reducing in-process tritium inventory from 2.03 kg to 0.48 kg (>4× reduction).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

FY22 Development of Improved Grout Waste Forms for Supplemental Low Activity Waste Treatment

About 54 to 56 million gallons of radioactive mixed waste is currently stored in underground tanks at the United States Department of Energy’s (DOE’s) Hanford site in the State of Washington. This waste will be separated into low- and high activity waste fractions, which will then be vitrified respectively into Immobilized Low Activity Waste (ILAW) and Immobilized High Level Waste (IHLW) products for subsequent disposal. The ILAW product will be disposed of in an engineered facility at the Hanford site while the IHLW product is designed for acceptance into a national deep geological disposal facility for high level nuclear waste. Treatment of the tank waste will take place in the Hanford Tank Waste Treatment and Immobilization Plant (WTP), which is under construction. However, since the WTP Low Activity Waste (LAW) Vitrification Facility was not designed to process the entire inventory of Hanford LAW, up to half of the retrieved Hanford LAW will require supplemental immobilization. Immobilizing LAW in a cementitious waste form known as Cast Stone has been investigated as a possible candidate supplemental immobilization technology

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Advanced Dimensional Depletion for Engineering of Reactors (ADDER) Software for Depletion and Fuel Management

The Advanced Dimensional Depletion for Engineering of Reactors (ADDER) software is being developed in the Research and Test Reactor (RTR) Program at Argonne National Laboratory to meet the reactor design and analysis needs of the Conversion Program. ADDER is a flexible tool that (1) provides a depletion capability through coupling external neutronics codes with a built-in CRAM solver or external depletion code and (2) provides a user-friendly interface to perform fuel management and criticality search operations. The ADDER software is a Python 3 application written using modern software development practices subject to a compliant implementation of NQA-1 and applicable Department of Energy software quality assurance standards. This report is the user guide for the software release referred to as ADDER v1.1.0. The motivation for a software to have flexible capabilities that ADDER possesses is the need to support a wide variety of geometries that are commonly required in analysis of research and test reactors. These reactors can have complex fuel, experiment, or control material shuffling patterns that persist over several years with many fuel management and partial refueling intervals. The scale of fuel management analysis can require tracking of an inventory that is multiple times the core loading. Many reactors, both power and non-power reactors of various types, will find the features of ADDER useful to facilitate key tasks that a fuel or core design engineer must perform with the convenience of concise input and validated functionality.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Vitrification Testing of HLW with High Phosphate

Projections of the number of high level waste (HLW) canisters to be produced in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) (e.g., [1]) are based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier Department of Energy (DOE) melter designs, incorporates a glass bubbler system. The bubblers create active glass pool mixing and thereby improve heat and mass transfer and glass melting rate. The WTP HLW melters each have a glass surface area of 3.75 m 2 and depth of ~1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by optimization of the feed and glass formulations, increasing the melter operating temperature above 1150⁰C, and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Life Cycle Analysis of Natural Gas Extraction and Power Generation: U.S. 2020 Emissions Profile

This analysis expands upon previous life cycle analyses (LCAs) of natural gas systems performed by the National Energy Technology Laboratory (NETL). It provides a complete inventory of emissions to air and water, water consumption, and land use change. These environmental burdens are detailed for all supply chain steps from natural gas production through natural gas distribution. This package includes the report, the NETL Natural Gas Model, and appendices that include several Excel workbooks and a python script to provide transparent access to the calculations and resulting data. This is revision 1 of the 2024 study (published December 17, 2024 and updated on January 24, 2025) and corrects a modeling error in natural gas composition. See the errata on page 2 for more information.

03 NATURAL GAS↗

Development of Decay Heat Sensitivity Analysis Capability in SCALE/ORIGEN

In this work, a decay heat sensitivity analysis capability was developed and implemented in the ORIGEN code of the SCALE nuclear modeling and simulation suite. This capability introduces improved numerical integration schemes, which overcome the challenges associated with accurately modeling the behavior of adjoint nuclide amounts during coarse time steps for both nuclide amount and decay heat sensitivity calculations. This capability significantly improves the accuracy of calculations without compromising computational efficiency compared to the existing method. Extensive verification was conducted for various benchmark problems, including a 238 Pu decay and an irradiation problem involving 135 Xe, evaluated with both coarse and fine time grids. The results show excellent agreement with reference direct perturbation solutions, reaffirming the computational accuracy of the newly proposed numerical integration methods. Furthermore, sensitivity analyses were performed for fission product inventories ( 147 Sm, 150 Sm, 155 Gd) in pressurized water reactor UO 2 and MOX fuel assemblies. These analyses demonstrated that the ORIGEN sensitivity analysis capability can capture detailed sensitivity coefficients and underlying physics in real applications. Additionally, a decay heat sensitivity analysis for high-assay low-enriched uranium fuel, including various initial 235 U enrichment and burnup points, highlights the extended capabilities of SCALE/ORIGEN in comprehensively assessing the factors influencing total decay heat. These advancements in ORIGEN offer valuable insights for reactor analysis, fuel design, and safety assessments, especially in the context of advanced nuclear fuel development and design changes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Hourly Load Profile Dataset for Federal, State, and Municipal Electric Vehicle Fleets in the United States

The electrification of U.S. federal, state, and municipal fleets is accelerating rapidly, driven by an increased availability of competitive electric vehicle (EV) options and supportive policies and targets. The dataset described in this report, accessible at data.nrel.gov/submissions/280, provides a critical foundation for identifying fleet electricity demand, projecting these future demands, and developing actionable strategies to support the widespread electrification of government fleets. The dataset incorporates available fleet data, including 54% of federal agency vehicles approved for analysis (notably, the U.S. Postal Service is absent). Additionally, it includes data from 50,000 state government vehicles and 94,000 local government vehicles. While this represents a small fraction of the 4.4 million vehicles owned by state and local governments reported by the Federal Highway Administration (2022), the framework supports future expansion as more fleet inventory data become available.

33 ADVANCED PROPULSION SYSTEMS↗

Leverage modern artificial intelligence (AI) enabled systems for waste reduction

Manufacturing industries continue to face challenges in reducing waste, as upstream strategies such as source reduction and product redesign require a deeper understanding of processes compared to conventional recycling methods. Recent advancements in artificial intelligence (AI) and machine learning (ML) have opened new opportunities to integrate modern computational techniques with traditional waste minimization strategies. This paper explores AI-enabled approaches for product redesign, source reduction, and recycling that can significantly reduce waste generation while improving efficiency and sustainability. AI-driven material substitution and lightweighting in product design enable discovery of novel materials with optimized properties, reducing waste without compromising performance. Reinforcement learning models optimize process parameters, raw material specifications, and machine sequencing to minimize production losses, while Industrial Internet of Things (IIoT) systems paired with AI analytics enhance real-time waste tracking, predictive maintenance, and quality inspection. Furthermore, AI-based demand forecasting and production planning reduce overproduction and excess inventory, as demonstrated in industrial applications. In recycling, ML-powered pattern recognition and robotic sorting technologies achieve higher accuracy in waste segregation, directly improving recycling efficiency. Complementary solutions such as smart bins and AI-enabled waste pickup scheduling optimize collection logistics, reducing both costs and emissions. Although implementation requires upfront investment in infrastructure and training, the long-term benefits include higher material efficiency, reduced waste, improved product quality, and stronger sustainability outcomes across the supply chain. By leveraging AI-enabled systems, manufacturers can align waste minimization efforts with circular economy principles, creating scalable solutions for both industry and society.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

Considerations for the Development of a Regenerable LANA.75 Bed for Tritium Storage

LaNi4.25Al0.75 or LANA.75 has been used by the Savannah River Site Tritium Facilities for decades to safely store hydrogen isotopes • High molar density at moderate pressures (~1000x that of gas at STP) = less glovebox space • Can be used as a pump by varying temperature = fewer moving parts • Can deliver He-3 free gas = fewer unit operations • Beds are limited life components due to accumulation of He-3 • Formation of “heel” • Inventory hold-up • Reduced capacity • Eventual weeping of He-3

STAACK, Gregory [Savannah River National Laborator↗

Reassessment of Dose Contributions from the M-Area Glass Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater (GW) and inadvertent human intruder (IHI) screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in U-235 being brought back into the list of radionuclides requiring inventory limits within the M-Area Glass SWF. In addition, while reviewing the earlier vadose zone (VZ) transport analysis efforts, an input deck error was observed in the U-234G VZ transport simulation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Natural Gas LCA

This presentation was provided via a pre-recorded presentation at the 2025 Coordinating Research Council Life Cycle Analysis of Transportation Workshop at Argonne National Laboratory on October 9, 2025. The presentation provides an overview of the methods and data sources used in the report, Life Cycle Analysis of Natural Gas Extraction and Power Generation: U.S. 2020 Emissions Profile and briefly discusses discrepancies between top-down and bottom-up inventory methods and how those might be overcome.

LCA↗

Expansion and Update of the R&D Feedstock Carbon Intensity Calculator (FD-CIC) for R&D GREET 2025

The R&D Feedstock Carbon Intensity Calculator (R&D FD-CIC) is a transparent and easy-touse tool to estimate feedstock-specific carbon intensity (CI) of biofuel feedstocks, and to examine potential CI variations of different farming practices to grow agricultural feedstocks for biofuel production. The objective of the tool is to assess potential effects of different farming practices with underlined factors to reduce uncertainties in such assessment and help R&D activities. The original work and early versions of R&D FD-CIC were supported by Advanced Research Projects Agency–Energy (ARPA-E) of the Department of Energy, with recent support by the Bioenergy Technologies Office of the Department of Energy. The tool uses life cycle inventory (LCI) data of key farming inputs from the R&D Greenhouse gases, Regulated Emissions, and Energy use in Technologies (R&D GREET) model. The system boundary of R&D FD-CIC covers the cradle-to-farm-gate activities, including upstream emissions related to farming input manufacturing and feedstock production.

09 BIOMASS FUELS↗

International Nuclear Security Nuclear Material Accounting and Control Kit Participant Guides

You work at a facility that stores large amounts of nuclear material in varying containerizations. Management has just approved the establishment of a tamper indicating device program in order to cut down on the effort required to conduct inventories, and to enhance overall control of nuclear materials within the site boundary. You have been assigned, along with your colleagues, to conduct an initial assessment to determine the ideal TID types to be used at your facility. You will apply TIDs to a variety of apparatuses that are representative of the closure mechanisms on different containers and entry points to rooms and vaults onsite.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Improved Fuel Particle and Compact Matrix Fission Product Release Prediction with Improved Mass Transfer Model

This study was motivated by the need to incorporate more physics-based mass transfer models in BISON in order to improve fuel particle and compact matrix fission product release prediction. This document details the integration of new modeling capabilities in BISON, including (1) development of sorption mass transfer models; improvement of thermal contact model by adding neon to the gas gap inventory, (3) implementation of PCEA and IG-110 graphite thermal models, (4) validation using Advanced Gas Reactor (AGR)-3/4 compact and capsule rings data. These new capabilities have been shown to enforce the desired physics with satisfactory accuracy. BISON’s predictions of fission product release of AGR-3/4 compacts compare favorably with PARFUME and the concentration profiles across capsule rings show reasonable trend to the experiment measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Analysis of Silver Release from Furnace-Tested TRISO Particles

High temperature testing of intact TRISO particles previously irradiated in the AGR-5/6/7 experiment was performed in the Furnace for Irradiated TRISO Testing (FITT) to directly confirm diffusive release of silver and europium from intact TRISO particles. Testing was conducted from 1,100–1,600°C for exposure times up to 100 h to directly confirm silver through layer release below safety testing temperatures. The FITT analysis showed highest levels of silver release at 1,300–1,400°C which confirms athermal release behaviors previous observed in step-wise and transient safety tests. Additionally, release was non-uniform with some particles releasing a majority of their inventory while others did not appear to release silver under identical testing conditions, which was consistent with historic observations. An assessment of the effective silver diffusion coefficient in the SiC layer was conducted and indicated maximum values in the 1,300–1,400°C range. The magnitude of the calculated diffusion coefficients also exceeded currently accepted diffusion coefficients. The release behavior of europium from intact particles was also analyzed in FITT for at 1,450–1,550°C for 500 h to 750 h. Direct confirmation of europium release below safety testing temperatures was confirmed absent contributions from matrix release. The analysis indicated europium release follows a general Arrhenius behavior and suggests general uniform release and indicates irradiation conditions influence observed release response. Calculated diffusion kinetics agree well with historic experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Industrial Electrification Assessment Framework

Electrification refers to the process of transitioning from fuel-powered systems to electric-powered systems, or electrotechnologies. This document first provides an overview of industrial electrification, including possible advantages and common applications. Next, it provides a comprehensive, step-by-step framework for assessing the potential for electrification at an individual industrial facility. The steps in this framework include inventorying current fuel-powered systems, evaluating a manufacturing facility’s electrification readiness, and identifying, evaluating, and prioritizing strategies and technologies. This document also contains information about project implementation, and the appendices provide tools and resources to support organizations pursuing electrification.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

SCALE Analyses of Scenarios in the Molten Salt Reactor Fuel Cycle

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, the capabilities of SCALE 6.3.1 for radionuclide characterization, criticality, and shielding were demonstrated through several scenarios in a molten salt reactor (MSR) nuclear fuel cycle. Three scenarios were selected for this study, with the molten salt breeder reactor (MSBR) serving as the reference design. The original thorium-based fuel salt used in MSBR was replaced with a 235 U-enriched fuel salt to better reflect anticipated future MSR concepts. The first scenario focused on the fuel salt preparation stage. Criticality analyses, using SCALE/CSAS-Shift and SCALE/SAMPLER, examined potential critical configurations in the fresh fuel salt container based on variables such as 235 U enrichment, UF 4 molar fraction, temperature, and container geometry. The second scenario explored the release of fission products during reactor operation in which the tritium buildup in the primary fuel salt was quantified through depletion calculations using SCALE/TRITON with continuous fuel makeup and fractional fission product removal. This scenario also assessed radioactivity levels in the off-gas system (OGS), quantifying radiotoxic nuclides that could be released into the environment. SCALE/SAMPLER was employed to perform uncertainty quantification to identify key operating parameters influencing the buildup of radiotoxic nuclides in the OGS. Additionally, radiation dose rates were calculated using SCALE/MAVRIC to estimate the maximum exposure levels within the reactor cell complex during operation. The final scenario evaluated the decay heat and radioactivity of the primary fuel salt collected in the drain tank, as well as the radiation dose rate outside the building from the presence of the drain tank. Overall, the application of SCALE’s various capabilities for radionuclide inventory generation, criticality analysis, and shielding in MSR fuel cycles was successfully demonstrated across these scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗