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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Radiation detector using a graphene amplifier layer

A radiation field is detected or imaged using one or more junction devices in which a two-dimensional conductor layer is capacitively coupled to a semiconductor absorber layer. In the junction devices, pixel-level amplification and read-out are accomplished through the photogating of the devices by absorption within the absorber layer while it is in a state of deep depletion. When the two-dimensional conductor is graphene, we refer to a device operating in that manner as a deeply depleted graphene-oxide-semiconductor (D 2 GOS) detector.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1991

The Cl-36 in the stratosphere

Initial measurements of the cosmogenic radionuclide, Cl-36, in the lower stratosphere were made by accelerator mass spectrometry. Samples were obtained using the large volume LASL air sampling pods on a NASA WB-57F aircraft. Untreated (for collection of particulates only) and tetrabutyl ammonium hydroxide treated (for collection of particulates and HCl) IPC-1478 filters were flown on three flights in the lower stratosphere. Chlorine (Cl) and Cl compounds are important trace constituents for stratospheric chemistry, in particular with respect to O3 destruction. Stratospheric Cl chemistry has recently received increased attention with the observation of strong O3 depletion in the Antarctic winter vortex and in the weaker and more complex Arctic winter vortices. Cosmogenic (Cl-36) is produced by spallation reactions from Ar mainly in the stratosphere, and has had several applications as a geochemical tracer. The large amounts of Cl-36 introduced by nuclear weapon testing have been removed from the stratosphere by now, and measurements in the stratosphere to obtain cosmogenic production rates and concentration distributions is now possible. The use of cosmogenic Cl-36 as a tracer for stratospheric Cl chemistry and for stratospheric/tropospheric exchange processes is investigated. A first attempt to determine stratospheric and tropospheric production rates, the partitioning of Cl-36 among particulate and gaseous Cl compounds, and the respective inventories and removal rates is being made. Results from a flight at 13.7 km, 30-33 degrees N, 97-107 degrees W, and from a second flight at 17.7 km, 43-45-36 degrees N, 92-94 degrees W, for the untreated and treated filters respectively are presented.

Deck, Bruce↗
DOE OSTI2025

Reduced-order modeling for efficient cross section library development in high-temperature gas reactor pebble-bed depletion analysis

Accurate modeling of running-in and equilibrium conditions in pebble-bed reactors (PBRs) requires precise microscopic multigroup neutron cross sections. In Griffin, deterministic neutronics calculations rely on multivariate interpolation over large cross section libraries, resulting in significant memory usage and performance bottlenecks. This work, together with a companion paper on Griffin integration, explores reduced-order models (ROMs) to replace interpolation with lightweight surrogates. Several ROM techniques are benchmarked, with deep neural networks (DNNs) demonstrating superior memory efficiency, scalability, and predictive accuracy. A total of 295 DNNs were trained to build a comprehensive isotope library, integrated into Griffin through a custom LibTorch interface for depletion analysis. Initial results demonstrate that DNN-based ROMs drastically reduce memory demands while preserving accuracy, enabling finer tabulations and additional state variables without overhead. In conclusion, the framework also supports online cross section generation and real-time DNN updates through transfer learning, improving fidelity by capturing self-shielding and evolving nuclide compositions during burnup.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Characteristics of U.S. Energy Production using Nuclear Fission

When considering potential energy production technologies for the future, a critical consideration centers on the question of “how green” the technology is. Here, the word “green” implies that the technology has a zero or minimal impact on the public and environment while still providing benefits by way of electricity, heat, and other products such as hydrogen and water. And, these potential impacts must be considered over the lifecycle of the technology deployment, from design, construction, operation, and disposition. Ideally, a green energy technology would be net-zero (i.e., having no to almost negligible contribution) on five key elements: 1. Greenhouse gas emissions, including carbon dioxide (CO 2 ), methane, nitrous oxide, and fluorinated gases such as ozone-depleting gases 2. Water consumption 3. Material resource consumption 4. Disposition of wastes 5. Public, flora, and fauna safety including deaths, health, and environmental impacts. Society would benefit from net-zero impacts of the five elements above while having low-cost energy. As society starts to replace fossil fuels in the energy mix, we need to consider the possible impacts of adopted technologies. We need a production approach that provides large quantities of energy while being safe, reliable, economical, and sustainable. In this report, we describe a variety of characteristics related to the five elements above to provide a fact-supported, science-based depiction of nuclear fission as an energy providing technology. By better understanding how fission power is nearly net-zero in the five elements above, we can position our thinking to align with the overarching goal to provide society with a low-impact energy source.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2014

The Nuclear Cryogenic Propulsion Stage

Nuclear Thermal Propulsion (NTP) development efforts in the United States have demonstrated the technical viability and performance potential of NTP systems. For example, Project Rover (1955 - 1973) completed 22 high power rocket reactor tests. Peak performances included operating at an average hydrogen exhaust temperature of 2550 K and a peak fuel power density of 5200 MW/m3 (Pewee test), operating at a thrust of 930 kN (Phoebus-2A test), and operating for 62.7 minutes in a single burn (NRX-A6 test). Results from Project Rover indicated that an NTP system with a high thrust-to-weight ratio and a specific impulse greater than 900 s would be feasible. Excellent results were also obtained by the former Soviet Union. Although historical programs had promising results, many factors would affect the development of a 21st century nuclear thermal rocket (NTR). Test facilities built in the US during Project Rover no longer exist. However, advances in analytical techniques, the ability to utilize or adapt existing facilities and infrastructure, and the ability to develop a limited number of new test facilities may enable affordable development, qualification, and utilization of a Nuclear Cryogenic Propulsion Stage (NCPS). Bead-loaded graphite fuel was utilized throughout the Rover/NERVA program, and coated graphite composite fuel (tested in the Nuclear Furnace) and cermet fuel both show potential for even higher performance than that demonstrated in the Rover/NERVA engine tests.. NASA's NCPS project was initiated in October, 2011, with the goal of assessing the affordability and viability of an NCPS. FY 2014 activities are focused on fabrication and test (non-nuclear) of both coated graphite composite fuel elements and cermet fuel elements. Additional activities include developing a pre-conceptual design of the NCPS stage and evaluating affordable strategies for NCPS development, qualification, and utilization. NCPS stage designs are focused on supporting human Mars missions. The NCPS is being designed to readily integrate with the Space Launch System (SLS). A wide range of strategies for enabling affordable NCPS development, qualification, and utilization should be considered. These include multiple test and demonstration strategies (both ground and in-space), multiple potential test sites, and multiple engine designs. Two potential NCPS fuels are currently under consideration - coated graphite composite fuel and tungsten cermet fuel. During 2014 a representative, partial length (approximately 16") coated graphite composite fuel element with prototypic depleted uranium loading is being fabricated at Oak Ridge National Laboratory (ORNL). In addition, a representative, partial length (approximately 16") cermet fuel element with prototypic depleted uranium loading is being fabricated at Marshall Space Flight Center (MSFC). During the development process small samples (approximately 3" length) will be tested in the Compact Fuel Element Environmental Tester (CFEET) at high temperature (approximately 2800 K) in a hydrogen environment to help ensure that basic fuel design and manufacturing process are adequate and have been performed correctly. Once designs and processes have been developed, longer fuel element segments will be fabricated and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREE) at high temperature (approximately 2800 K) and in flowing hydrogen.

Houts, Michael G.↗
DOE OSTI2024

Spatiotemporal characterization of cerium monoxide in laser ablation plasmas using spectrally-resolved fast-gated imaging

The impact of oxidation chemistry on the emission characteristics and spatial structure of laser-produced Ce metal plasmas was investigated using laser-induced breakdown spectroscopy (LIBS) and time-resolved fast-gated imaging employing narrowband optical filters. Images of the plasma emission show that CeO coexists with atomic species in the periphery and vortex ring of the plasma plume. Image processing was also applied to combine independent monochromatic images of the plasma emission into a single merged image, culminating in a timelapse on the spatiotemporal evolution of atomic and molecular species within the plasma plume. The formation of CeO species was observed to proceed faster for plasmas generated in atmospheres containing larger concentrations of oxygen based on ratios of CeO-to-atomic emission intensities. Furthermore, these same ratios were shown to plateau and decrease at later times (≥25 μs), potentially indicating the depletion of CeO number densities in the plasma as the monoxide undergoes reactions to form higher polyatomic oxides. Altogether, these results provide fundamental insights into the chemical dynamics and intermixing between plasma-gas species in laser ablation cerium plasmas, advancing our understanding on optical signatures of nuclear-relevant materials to enable in-field measurement capabilities.

47 OTHER INSTRUMENTATION↗
NASA NTRS1973

Induction simulation of gas core nuclear engine

The design, construction and operation of an induction heated plasma device known as a combined principles simulator is discussed. This device incorporates the major design features of the gas core nuclear rocket engine such as solid feed, propellant seeding, propellant injection through the walls, and a transpiration cooled, choked flow nozzle. Both argon and nitrogen were used as propellant simulating material, and sodium was used for fuel simulating material. In addition, a number of experiments were conducted utilizing depleted uranium as the fuel. The test program revealed that satisfactory operation of this device can be accomplished over a range of operating conditions and provided additional data to confirm the validity of the gas core concept.

Poole, J. W.↗
DOE OSTI2022

Forward silicon vertex/tracking detector design and R$\&$D for the future Electron-Ion Collider

The proposed high-luminosity high-energy Electron-Ion Collider (EIC) will provide a clean environment to precisely study several fundamental questions in the fields of high-energy and nuclear physics . A low material budget and high granularity silicon vertex/tracking detector is critical to carry out a series of hadron and jet measurements at the future EIC especially for the heavy flavor product reconstruction or tagging. The conceptual design of a proposed forward silicon tracking detector with the pseudorapidity coverage from 1.2 to 3.5 has been developed in integration with different magnet options and the other EIC detector sub-systems. The tracking performance of this detector enables precise heavy flavor hadron and jet measurements in the hadron beam going direction. The detector R&D for the proposed silicon technology candidates: Low Gain Avalanche Diode (LGAD) and radiation hard depleted Monolithic Active Pixel Sensor (MALTA), which can provide good spatial and timing resolutions, is underway. Bench test results of the LGAD and MALTA prototype sensors will be discussed.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

Reaction Rate Ratios for Recent Fast Metal Experiments with Large Plutonium Masses

Reaction rate ratios are integral responses that are used within the criticality experiments field because they contain spectral information. While these types of measurements have been utilized for nuclear data validation with historic experiments, few experiments of this type have been utilized for recent experiments, as few exist. This work focuses on measured reaction rate ratios for two nearly bare plutonium critical assemblies with different geometries: one that is cube like (with a Pu mass of 40 kg) and one that is slab like (with a Pu mass of 109 kg). Irradiations were performed with both configurations in which foils were placed near the center of the assembly. Plutonium, highly enriched uranium, depleted uranium, and Au foils were included in the irradiation and counted via high-purity germanium detectors. From these measurements, reaction rate ratios were calculated. Measured and simulated values and uncertainties are presented for the reaction rate ratios. Ratios utilizing the following reactions are given in this work: 197 Au(n, γ), 197 Au(n,2n), 235 U(n,fission), 238 U(n, fission), 238 U(n,2n), 238 U(n,γ), and 239 Pu(n,fission). Uncertainties for the measured reaction rate ratios ranged from 4% to 7%, and the contribution of various parameters to this uncertainty was investigated. The results are compared to historical experiments and should be used for nuclear data validation for future nuclear data library releases. These measurements are part of the EUCLID (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) project, which utilizes measurement responses in addition to k eff (such as these reaction rate ratios) to help reduce uncertainties in 239 Pu nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Initial Engineering Framework for Species Tracking in MSRs Involving Fuel Salt and Structures

This report implements a high-fidelity multiphysics modeling framework using the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program tools to track leaching and plating of materials between the fuel salt and structures in Molten Salt Reactors (MSRs). Specifically, this framework is demonstrated in modeling various processes within MSRs including hot and cold leg corrosion, redox potential driven corrosion, colloid precipitation and deposition, and noble metal plating and decay heat modeling. The model integrates neutronics, thermal-hydraulics, depletion, and thermochemistry to simulate the evolving chemical behavior of the fuel salt in interactions with structural materials. Initial findings suggest that hot-cold leg corrosion in MSR systems may become significant if the redox potential of the fuel salt is not sufficiently controlled. Additionally, the impact of noble metal decay heat deposition in large power reactors is explored. The resulting engineering framework for species tracking will be applied in the future toward specific validation work with ongoing experimental efforts focusing on both chloride and fluoride salt loop experiments with in-situ corrosion and redox control instrumentation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Tribocorrosion of stainless steel sliding against graphite in FLiNaK molten salt

The molten salt reactor (MSR) concept recently gained renewed interest in developing Generation IV nuclear reactors. One MSR design uses fluoride molten salts to cool tri-structural isotropic particle fuel encapsulated by a carbon matrix into spherical pebbles, which would inevitably contact the reactor's stainless steel container wall during salt circulation. Characterizing this interaction is crucial for reactor safety. Here, in this work, we report the tribocorrosion behavior of graphite sliding against Type 316H stainless steel lubricated by a FLiNaK (LiF:NaF:KF; 46.5:11.5:42 mol %) molten salt in an argon environment. Accelerated wear loss was observed at a higher temperature because of a lower molten salt viscosity and a higher corrosion rate. The graphite had a more rapid material loss at a higher sliding speed than stainless steel because of its higher vulnerability to vibration-induced microfracture. The salt-starved condition caused more material loss than either the no-salt or the salt-flooded condition because neither a graphite transfer film nor stable boundary lubrication could be established at salt starvation. An interesting dual-layer surface film was discovered on the stainless steel worn surface: a chromium-rich top film and a nickel-accumulated but chromium-depleted interlayer. The film composition and structure provide fundamental insights to the mechanochemical interactions between stainless steel and graphite in a molten salt environment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Transition Core Modeling for Extended Enrichment & Accident-Tolerant Fuels Using Polaris/PARCS

Commercial light water reactor (LWR) operators and fuel vendors are currently interested in increasing the low-enriched uranium (LEU) fuel enrichments from the current limit of 5.0 $^w/_o$ $^{235}U$ up to 10 $^ w/_o$ $^{235}U$ (referred to as "LEU+") in their current fleets; they are also interested in using accident-tolerant fuel (ATF) with both LEU and LEU+ fuel. This report aims to identify modeling challenges and accuracy concerns in transition core analysis using the SCALE Polaris lattice physics code and U.S. Nuclear Regulatory Commission core simulator PARCS. At the time this study was started, no publicly available LEU+ core designs existed for boiling water reactor (BWR) or pressurized water reactor (PWR) systems. Therefore, fuel lattices were shuffled within a multi-assembly model to mimic neutronically challenging lattice combinations seen in transition cores, such as a fresh LEU+ lattice next to depleted LEU lattices. In addition to multi-assembly models, whole-core BWR transition core calculations were performed for ATF and LEU+ fuel using an existing Hatch-1 Cycle 3 core model. A whole-core BWR model was chosen due to the more heterogeneous core designs compared to those for a PWR core. Since the original core is an old checkerboard core design and no core or fuel design optimization was performed for the modeled fuel types, these core calculations were intended only to provide: (1) Comparisons of core characteristics of interest, such as the pin power distributions and peaking factors, Doppler temperature coefficients (DTCs), and control blade worths (CBWs) under challenging core designs, (2) Identification of reactor physics challenges in modeling LEU+ and ATF cores, and (3) A stress test for the Polaris/PARCS two-step modeling approach, including characterization of the relative accuracy for predicting characteristics of interest such as pin power distributions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Milestone 1.2.10: Steady-state H 2 “roll over” point data for aluminum alloys 1100 and 6061

Extended (> 50 years) dry storage is being evaluated by the U.S. Department of Energy (DOE) for the disposition of ~ 18 metric tons of aluminum-clad spent nuclear fuel (ASNF). Transition of the current ASNF inventory into dry storage—using the standard DOE canister—necessitates a rigorous, predictive understanding of the long-term physical and chemical factors that may influence the integrity of the proposed storage canister, including radiolytic molecular hydrogen (H 2 ) generation. Current model predictions employ initial radiolytic yields of H 2 , the values of which change as the cladding’s H 2 -precursor inventory is depleted and H 2 itself becomes progressively more involved in radiolytic and surface dissociation processes. Consequently, the absorbed radiation dose that this steady-state H 2 yield corresponds to is essential for the evaluation and improvement of model predictions. Here, we report our findings on the long-term generation of H 2 from the gamma irradiation (≤ 36 MGy) of corroded AA1100 and AA6061 coupons in helium environments at ambient temperature and ~ 50% RH. Our findings show that AA1100 systems reached steady-state by ~ 36 MGy, while higher doses were necessary for AA6061 systems. This discrepancy was attributed to the AA6061 coupons developing a thicker corrosion layer that led to the trapping of H 2 and its precursors, and potentially additional chemistries, ultimately delaying the depletion of H 2 precursors and the system’s “roll over” point. Further, current model predictions—based on previous AA1100 data—do not show steady-state attainment until above 120 MGy, which is not the case for the AA1100 data collected here. Consequently, the new alloy dependent data presented here are important for the continued improvement of predictive computer models for evaluating the feasibility of extended storage of ASNF in helium backfilled canisters.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Isotopic and Fuel Lattice Parameter Trends in Extended Enrichment and Higher Burnup LWR Fuel, Volume II: BWR

Commercial light water reactor (LWR) operators and fuel vendors in the United States are pursuing changes to nuclear fuel that include extended enrichment (EE) and accident-tolerant fuel (ATF) designs to further improve reactor safety and plant economics. Extended fuel enrichments above 5% 235 U pin enrichment and up to 10% 235 U are a subset of high assay low-enriched uranium (HALEU) that may be deployed in commercial US LWRs in the near term. ATF features, such as cladding coatings or alternative cladding materials, are designed to improve fuel system performance under accident conditions. One goal of EE is to improve fuel cycle economy by enabling fuel to be depleted to higher burnup than the typical current limits (62 gigawatt-days per metric ton of uranium [GWd/MTU]). Adoption of EE, ATF, and high burnup (HBU) fuels in the US commercial fleet requires a clear understanding of the effects on core physics parameters and used fuel isotopic content, as well as confidence in the accuracy of computer code predictions over an expanded range of materials, enrichment, and burnup. A thorough understanding of the applicability and adequacy of benchmark data (e.g., criticality, decay heat, isotopic content) for computer code validation is necessary to ensure that appropriate safety margins are maintained. As part of the US Nuclear Regulatory Commission (NRC) agreement number 31310019N0008, “SCALE Code Development, Assessment and Maintenance,” the effects of EE, ATF, and HBU are being assessed for selected representative LWR fuel designs. The project is divided into phases, and this report summarizes the findings of the Phase 1 work, which focuses on lattice physics parameter and used fuel isotopic changes for a conventional GE14 10 x 10 boiling water reactor (BWR) design with GNF-2 part length rod patterns to model a modern BWR assembly design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Modeling a Thermal-Spectrum LEU-fueled Molten Salt Reactor Co-fueled with Thorium in SCALE 6.3.1 and Serpent-2

Interest in the development of advanced nuclear fission reactors for commercial electricity production has risen in recent years, with Generation IV reactor designs offering numerous advantages in safety, efficiency, fuel cycle sustainability, and waste management. In particular, thorium-fueled molten salt reactors (MSRs) are considerably promising for enhancing fuel cycle sustainability in their ability to breed fissile 233U fuel from thorium, a presently untapped and widely abundant resource. A novel MSR fuel cycle concept, the ”Sourdough” refueling and waste management strategy, has previously been demonstrated with a traditional uranium-based fuel cycle in a thermal-spectrum MSR operating with low enriched uranium (LEU) fuel with favorable neutronic performance. However, the ability to use this unique fuel cycle approach with thorium-based molten salt fuels has not yet been studied. In this work, the Sourdough fuel cycle was implemented in a small, thermal-spectrum MSR fueled with high assay low enriched uranium (HALEU) and fertile 232Th for breeding 233U fuel. Relevant neutronic data, including fuel and isothermal temperature feedback behavior, was studied using the SCALE 6.3.1 and Serpent-2 code systems, and keff data was measured during simulated depletion at 400 MWth. The Sourdough fuel cycle concept is shown to perform favorably with a thorium-fueled MSR model, thus warranting further study into its use in other MSR designs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of a high-fidelity multi-physics coupling between MCNP6.2 and CTF4.0 for VVER applications

Ensuring system safety in the design, licensing, and operation phases is a priority in the nuclear industry. Performing extensive, full-scale reactor safety experiments is often prohibitive due to the large associated costs. Computational simulations offer an alternative safety analysis method, typically with significant cost reductions. Recent high-level developments in technology and increased availability of computational resources have allowed the development of high-fidelity, high-resolution multi-physics coupled codes. Such developments may be used to generate reference models for deterministic core calculations. Under this framework, the high-fidelity continuous energy Monte Carlo-based neutron transport code MCNP6.2 was coupled externally with the state-of-the-art thermal-hydraulics subchannel code, CTF4.0 for VVER (Water-Water Energetic Reactor) applications. A VVER-1000 fuel assembly model was used to demonstrate the capability of the coupled code. The converged coupled solution is compared to initial results, consisting of the first MCNP evaluation and first CTF evaluation after initialization. The VVER-1000 type assembly results are compared to other evaluations of the same assembly model. The findings indicate good agreement with expectations and reference cases, where available. The initial results of the coupled MCNP6.2/CTF4.0 calculations at the assembly level for steady-state calculations are presented in this study, which may support future work toward high-fidelity coupled full-core modeling capabilities. The multi-physics model may be further improved, enhanced, and expanded for both cycle depletion and transient applications. Such a multi-physics system will also be applicable to the VVER-1200 and other triangular lattice designs and support their deployment and operation safely and economically. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Accelerating effects of galvanic corrosion and dissimilar materials on the corrosion of 316H in NaCl-MgCl2 salt

Corrosion of materials presents a significant challenge for the long-term operation of molten salt reactors. This study aims to identify the most effective techniques for evaluating the corrosion performance of materials in molten salts, with a focus on the effects of galvanic corrosion and dissimilar materials. A reliable testing methodology for assessing material corrosion in molten chloride salts has been successfully developed. The corrosion of Alloy 316H in molten NaCl-MgCl2 salt was found to be significantly accelerated by galvanic corrosion. Additionally, the presence of dissimilar materials resulted in a slight increase in the corrosion rate of Alloy 316H in NaCl-MgCl2 salt. Microstructural characterization was utilized to understand the corrosion behavior of test samples under different conditions. Common trends observed across samples include chromium depletion and iron enrichment near corroded surfaces. Molybdenum enrichment along grain boundaries and corrosion surfaces was also frequently noted.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Design and construction of a new detector to measure ultra-low radioactive-isotope contamination of argon

Large liquid argon detectors offer one of the best avenues for the detection of galactic weakly interacting massive particles (WIMPs) via their scattering on atomic nuclei. The liquid argon target allows exquisite discrimination between nuclear and electron recoil signals via pulse-shape discrimination of the scintillation signals. Atmospheric argon (AAr), however, has a naturally occurring radioactive isotope, 39Ar, a ß emitter of cosmogenic origin. For large detectors, the atmospheric 39Ar activity poses pile-up concerns. The use of argon extracted from underground wells, deprived of 39Ar, is key to the physics potential of these experiments. The DarkSide-20k dark matter search experiment will operate a dual-phase time projection chamber with 50 tonnes of radio-pure underground argon (UAr), that was shown to be depleted of 39Ar with respect to AAr by a factor larger than 1400. Assessing the 39Ar content of the UAr during extraction is crucial for the success of DarkSide-20k, as well as for future experiments of the Global Argon Dark Matter Collaboration (GADMC). This will be carried out by the DArT in ArDM experiment, a small chamber made with extremely radio-pure materials that will be placed at the centre of the ArDM detector, in the Canfranc Underground Laboratory (LSC) in Spain. The ArDM LAr volume acts as an active veto for background radioactivity, mostly ?-rays from the ArDM detector materials and the surrounding rock. This article describes the DArT in ArDM project, including the chamber design and construction, and reviews the background required to achieve the expected performance of the detector.

Noble liquid detectors, Dark Matter detectors, Cry↗