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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Study of Applications of Radiological Data Fusion for Nuclear Security and Non-proliferation Applications

This report describes the radiological Scene Data Fusion (SDF) technology and the potential to apply the SDF technology to a variety of problems related to nuclear non-proliferation, radiological security, and other missions related to the national nuclear security agency and the US government, more broadly. The application space was conceived by both the authors of this report and the many scientists and stakeholders they interviewed. In order to successfully apply the SDF technology to the various applications, further research and development is often necessary and those R&D steps are also outlined herein.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Survey of emerging nuclear data needs for nonproliferation applications with advanced reactors

Nuclear science plays a key role in non-proliferation activities supporting advanced reactor technologies. Nuclear data underpin predictions and interpretations of nuclear material behavior and signatures in reactor fuel production, use, transport, and storage. Advanced reactors provide new challenges compared to the current fleet of thermal fission reactors. This report consolidates reported nuclear data needs from representative workshops, conferences, and publications, identifying six themes for recommended future investments supporting non-proliferation and safeguards applications. While also identified as data needs, major fission product evaluations and (α,n) reactions were omitted as there are ongoing activities producing new data under NA22/Objective O. Each theme is summarized below with example data and association with the nonproliferation mission for advanced fuels and reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Knowledge Transfer Program for International Nuclear Safeguards

Nuclear safeguards were first announced in 1945 by the U.S., Canadian and British governments as a means to exchange scientific information about peaceful uses of atomic energy and prevent the use of nuclear material for weapons. International nuclear safeguards, now under the provision of the International Atomic Energy Agency (IAEA), serves to hold accountable the 140 member countries (States) that have entered into treaties and agreements against the spread of nuclear weapons. Most notably, the IAEA acts as a nuclear materials inspectorate under the global Nuclear Non-Proliferation Treaty, brought about in 1968, that specifies commitments member States make to the world?s non-proliferation regime. The Office of International Nuclear Safeguards (OINS) is part of the National Nuclear Security Administration, a semi-autonomous agency within the U.S. Department of Energy. Within OINS, the Human Capital Development program recognizes the need to build workforce capacity and safeguards expertise. This includes supporting the education and training of younger generations working in international nuclear safeguards. One of the program?s main concerns is core competencies being lost to retirement or attrition and building knowledge retention pipelines from senior to new professionals. One identified core competency is knowledge gained by individuals who have completed IAEA assignments such as missions to member States to conduct safeguards inspections. Such individuals possess unique knowledge, and efforts are being put forth to effectively capture this knowledge within the national laboratories complex. To this end, we report findings from a one-year mentor-mentee knowledge retention program in which an international safeguards subject matter expert imparted knowledge and skills to a willing professional mentee. The knowledge and skills stem from the mentor?s multi-year assignment at the IAEA in Vienna, Austria.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
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 OSTI2025

Demonstration of active neutron interrogation of special nuclear materials using a high-intensity short-pulse-laser-driven neutron source

Detecting shielded special nuclear material, such as nuclear explosives, is a difficult challenge pursued by non-proliferation, anti-terrorism, and nuclear security programs worldwide. Interrogation with intense fast-neutron pulses is a promising method to characterize concealed nuclear material rapidly but is limited by suitable source availability and proven instrumentation. In this study we have pioneered a demonstration of such an interrogation method using a high-intensity, short-pulse, laser-driven neutron source that offers potential benefits compared to conventional neutron sources. The measurement results reported here represent the first experimental demonstration of this interrogation approach on enriched uranium items and demonstrate the feasibility of a precise measurement using realistic nuclear materials, representative of field scenarios, even with just a single laser-driven neutron pulse. Bright pulsed sources can overcome the nuisance background of items with strong internal neutron sources, improving analytical power, while single-shot assay is attractive in high-throughput situations where time is at a premium. The science and technology of this type of neutron production is developing rapidly, and we anticipate that practical mobile interrogation systems will become available based on the detection concepts demonstrated here to meet the growing measurement needs.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

FAST Irradiations, Postirradiation Examinations, and Modeling of U-Mo for Light Water Reactor Applications

Many next generation light water reactor (LWR) concepts, such as mobile small modular reactors, are seeking to use smaller core dimensions than conventional reactor types. Smaller reactor cores require an increase in fissile material to maintain reactivity. For non-proliferation purposes, enrichment increases are limited to less than 20% (high assay low enriched uranium, [HALEU]) and so higher uranium density fuels than UO 2 must be considered. To this end, uranium-molybdenum alloys were tested using the Fission Accelerated Steady-state Test (FAST) approach. The experiment test matrix is focused on identifying the temperature transition between low swelling and high fission gas retention to break away swelling and low fission gas retention. This paper documents the results of irradiation tests and post-irradiation examinations (PIE) including neutron radiography, rodlet profilometry, fission gas collection analysis, and optical metallography. The results of these tests showed that unconstrained U-Mo fuels (solid, Na-bonded rodlets) have a swelling threshold between 400-450°C with minimal fission gas release below this point. Higher temperature solid fuel showed microstructural zoning with small pore networks while lower temperature solid fuels have a uniform microstructure with large pore networks. U-Annular Mo fuels where swelling had some self-constraint imposed upon it, were shown to have much reduced swelling compared to their solid counterparts as well as very low fission gas release for irradiation temperatures up to 500°C. These initial results show that the use of U-Mo in constrained fuel geometries could be used as a high uranium density HALEU fuel for LWRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Advancing International Integration and Strengthening Responsible Peaceful Uses of Nuclear Applications Through Specialized Curriculums

Nuclear technology has been pivotal in addressing some of the most pressing global challenges, ranging from energy production to combatting infectious diseases to agricultural security. Oak Ridge National Laboratory (ORNL), as a leader in nuclear research and development including in the field of radioisotopes, is well situated to share lessons learned and enhance global collaboration from years of discoveries in the field. Accordingly, the U.S. Department of Energy’s National Nuclear Security Administration (NNSA) sponsors specialized educational programs at ORNL, with support from the IAEA, focused on advancing peaceful nuclear applications and associated industries while upholding strong nuclear safety, security, and safeguards standards. The Joint U.S./IAEA International School on Peaceful Uses of Nuclear Applications, launched in 2024, is a cornerstone of this collaboration. The school provides an opportunity for early-career professionals from around the globe to gain practical knowledge and skills in utilizing nuclear technologies for peaceful purposes. This initiative demonstrates the United States’ commitment to its obligations under the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), specifically Article IV, which calls on nuclear-weapon states to facilitate access to the peaceful uses of nuclear energy while guarding against the proliferation of nuclear weapons. In this paper, we explore the curriculum, objectives, and global impact of the program. Participants of the ICARST-2025 conference are invited to learn more about this program, contribute to its development, and explore opportunities for collaboration. This school exemplifies how strategic partnerships, and educational initiatives can drive the peaceful and beneficial use of nuclear technology worldwide

Raffo Caiado, Ana [ORNL] (ORCID:0009000239304805)↗
DOE OSTI2025

Application of a Triple Bubbler Inside Molten Salt for Pyroprocessing Safeguards Purposes

The application of a triple bubbler system within molten salt environments, specifically a eutectic mixture of LiCl-KCl, is explored for nuclear safeguards purposes, focusing on nuclear material accountancy (NMA) and holdup accountancy in pyroprocessing techniques. Traditional holdup accountancy methods, which involve periodic sampling and lengthy lead times, are being supplemented by the development of actinide sensors using electrochemistry. However, these methods alone are insufficient to meet regulatory requirements set by agencies like NRC and IAEA. The triple bubbler device measures salt level, density, and surface tension simultaneously using three dip tubes of varying diameters and heights. This study investigates the accuracy and robustness of the triple bubbler in molten salt conditions at temperatures ranging from 450-550°C. The updated bubbler design, utilizing tantalum metal and larger tube geometry, demonstrated minimal plugging and maintained measurement accuracy within acceptable uncertainties: 0.611% for depth, 1.34% for density, and approximately 40% for surface tension. Results indicate that the triple bubbler system is effective for real-time, accurate measurements in pyroprocessing environments, making it a promising tool for ensuring compliance with nuclear safeguards and supporting nuclear non-proliferation efforts.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A Mobile Hot Cell for Conditioning Disused Sealed Radioactive Sources for Storage or Transportation

An innovative Mobile Hot Cell (MHC) has been developed for conditioning Disused Sealed Radioactive Sources (DSRS) category 1 and 2 for storage or transportation. The MHC is designed to provide both Radiological and Biological containment with a maximum capacity of 1000 Ci Co60 or 5000Ci Cs137 source and can be transported via standard cargo containers. This project has been supported through the National Nuclear Safety Administration (NNSA) Offsite Source Recovery Program (OSRP). The project is intended for the international community rather than domestic although domestic use is a possibility. Many countries have significant stockpiles of these devices that are often stored in less-than-optimal circumstances. This necessitates that these devices be addressed expeditiously, and the sources secured. The MHC utilizes robotics, automation, and other non-traditional methods for disassembling, characterizing, and packaging these sources that have reached end of life or are otherwise not needed. These innovative approaches are necessary to facilitate an expedited timeline to efficiently and safely secure these sources in a non-proliferation effort. Conditioning efforts include disassembling the device such as a teletherapy head used for cancer treatment, or blood/research irradiators such that the radioactive sources may be removed safely. The sources are then characterized. Leak checks are performed, dimensions are verified, and serial numbers are confirmed. Upon completion, the sources are typically placed into a Standard Forms Capsule which is seal welded closed. It is leak tested and placed into a Long-Term Storage Shield (LTSS) which can either be secured for storage directly or loaded into an appropriate cask for transportation. Further innovations include multiple deployment scenarios that include a full deployment of MHC components, deployment of the MHC automation internal components to an existing hot cell, deployment of minimally required MHC components and incorporation of sand for shielding, and integration of the MHC for Silo Storage, or Bore Hole Storage efforts. The MHC has evolved from a very specific use case to a “Swiss Army Knife” type of a tool in that it can be readily adapted to a large variety of situations. Innovative approaches such as the use of robotics, Computer Numeric Control (CNC) machining centers, automated welding equipment, HDMI Cameras, and LED lighting are some of the developed technologies incorporated into the MHC design. Shielding is accomplished with a steel walled Base Box which is surrounded by four nesting doll shield shells which when combined limits the external dose rate to 5mr/hr when a 1000 Ci Co60 source is exposed inside.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Evaluation of additive manufacturing side channels for nuclear nonproliferation applications

The nuclear industry is continuing to grow and adopt new technologies to manufacture components. One of the technologies under consideration is additive manufacturing (AM). AM could make production of components at lower cost and could lessen lead times significantly. However, AM could be prone to proliferation risks. AM machines give off signatures during the manufacturing process, and these signatures offer an alternative means of observing the component. Correlations between these signatures and the geometry being manufactured could be developed; enabling proliferators as they could take advantage of these correlations to steal AM manufacturing instructions relating to nuclear technology. This work pertains to preliminary work exploring the feasibility of tapping into AM side channels to predict geometries being manufactured on AM machines. The feasibility of utilizing vibration to draw geometric correlations will be assessed on an Ultimaker 2 + thermoplastic 3D printer.

Additive manufacturing↗
DOE OSTI2025

Assessment of Low-Level Pu Isotope Ratio Measurements Using Multicollector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS/MS) Equipped with a Pre-Mass Filter

We present an initial investigation into the performance of a multicollector inductively coupled plasmamass spectrometer equipped with a pre-mass filter (Neoma MC-ICP-MS/MS) for making plutonium (Pu) isotope ratio measurements on solutions containing low level (i.e., pg mL –1 ) Pu concentrations. This assessment was achieved by comparison of the 240 Pu/ 239 Pu, 241 Pu/ 239 Pu, and 242 Pu/ 239 Pu ratios attained over a one month period on the MC-ICP-MS/MS with the long-term (∼1 year) performance observed on the predecessor MC-ICP-MS (Neptune Plus) instrument each equipped with an equipped with an APEXΩ desolvating nebulizer for repeated measurements of certified reference materials from New Brunswick Program Office (NBL PO) CRM 136a and CRM 137. The MC-ICP-MS/MS performance of repeated measurement of CRM 136a (n = 20) resulted in mean values of 240 Pu/ 239 Pu = 0.1448 ± 0.0006, 241 Pu/ 239 Pu = 0.00371 ± 0.00006, and 242 Pu/ 239 Pu = 0.00682 ± 0.00006 (k = 2). The CRM 137 (n = 20), analyzed during the same analytical sessions, produced mean values for 240 Pu/ 239 Pu = 0.2414 ± 0.0006, 241 Pu/ 239 Pu = 0.00464 ± 0.00007, and 242 Pu/ 239 Pu = 0.0157 ± 0.0001 (k = 2). These results closely align with the certificate values for CRM 136a and CRM 137 and are within the k = 2 envelopes defined by the long-term performance of the traditional MC-ICP-MS approach (Neptune Plus). Examination of the performance of the various Pu isotope ratios as a function of total Pu content revealed accurate results (<3% relative difference, or RD) above ∼50 fg total Pu. The results presented here demonstrate the capability of the MC-ICP-MS/MS making accurate and precise low level Pu isotopic measurements. While the intent of this work was not to investigate the functionality of the collision cell, the pre-mass filter was employed. Future studies are warranted to investigate the entire capability of the MC-ICP-MS/MS collision cell and pre-cell mass filter optimization for performing low level Pu isotope measurements, even in mixed matrix samples.

CRM↗
DOE OSTI2025

Portal Monitoring Considerations for Treaty Verification Applications

Portal monitoring is one method in a suite of options for treaty verification and can assist in verification of treaty accountable items (TAIs). Portal monitor technology spans a wide range of sensors, including radiation detectors, break-beams, or weight sensors. For radiation-detecting portal monitors, they are able to confirm absence / presence of a radiological signature and to track TAI direction of motion. Portal monitoring uses measurements from strategically placed sensors (e.g., radiation detectors) to record the entry or egress of TAIs. The implementation of portal monitoring relies on three key elements: perimeter definition, portal locations, and sensor technology, and can be implemented in either time-bound or continuous operations depending on what is permitted by the Treaty.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Assessing the Heat Transfer Modeling Capabilities of CFD Software for Involute-Shaped Plate Research Reactors

The ongoing efforts to convert High-Performance Research Reactors (HPRRs) using Highly Enriched Uranium (HEU) to Low-Enriched Uranium (LEU) fuel require reliable thermal–hydraulic assessments of modified core designs. The involute-shaped fuel plates used in several major HPRRs present unique modeling challenges due to their compact core geometries and high heat flux conditions. This study evaluates the capability of three commercial CFD tools, STAR-CCM+, COMSOL, and ANSYS CFX, to predict cladding-to-coolant heat transfer using Reynolds-Averaged Navier–Stokes (RANS) methods within the thermal–hydraulic regimes of involute-shaped plate reactors. Broad sensitivity analysis was conducted across a range of reactor-relevant parameters using two turbulence models (k−ϵ and k−ω SST) and different near-wall treatment strategies. The results were benchmarked against the Sieder–Tate correlation and experimental data from historic studies. The codes produced consistent results, showing good agreement with the empirical correlation of Sieder–Tate and the experimental measurements. The findings support the use of these commercial CFD codes as effective tools for assessing the thermal–hydraulic performance of involute-shaped plate HPRRs and guide future LEU core development.

CFD↗