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At least 523 records · Page 29

Thermal Scattering Law Data Development for Paraffin Wax

Paraffin wax is often used as a nuclear moderator to slow down the fast neutrons in experimental critical assemblies [1]. It is a colorless and soft solid material that consists primarily of straight-chain alkanes (n-alkanes), which are hydrocarbons with the general formula CnH2n+2 [2-3]. The length of the hydrocarbon chain ranges from C20 to C30 and higher [2]. It is distinguished by its solid state at room temperature and begins to melt above approximately 310 K [4]. Paraffin wax is a commonly employed substance in the manufacture of shielding. One of its noteworthy characteristics is its ability to effectively absorb the neutrons. Also, it possesses a high macroscopic cross section, which enables it to efficiently moderate neutrons. As a result, paraffin wax is extensively utilized in various applications where moderation and shielding of neutrons are needed. For simulations, it is necessary to evaluate its thermal scattering law (TSL) and cross sections. Computationally, classical molecular dynamics (CMD) simulations provide the capability of simulating atomic details. For example, several unary, binary, and few multi component mixtures have been investigated of the paraffin model by using molecular dynamics simulations [5-12]. An assessment of thermal neutron scattering in a heavy paraffinic oil treated both as a solid and a viscous fluid containing 25% linear branched paraffin (C30H62), 35% one ring cycloalkane (C30H60), 15% two rings cycloalkane (C30H58), and 25% aromatic (C30H60) chains has been studied using CMD simulations for producing TSL data [13]. Nevertheless, there is lack of TSL and cross section data for paraffin wax as most of the reported analyses focus on the unary and binary mixture of n-alkanes, which is not consistent with actual paraffin wax [2]. In this work, we applied the equilibrium CMD simulations technique to explore the structure and dynamical properties of wax, which are fundamental input to calculate the TSL. A paraffin wax system was modeled using the CMD code LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator) [14-15] with the semi-empirical COMPASS [16] force field. The density of state (DOS) was calculated from the normalized velocity autocorrelation function (VACF), which is the Fourier transform of the normalized VACF. The DOS was used for the calculation of the TSL and thermal scattering cross sections. The paraffin wax atomic system was constructed by using the MedeA material design platform [17], and was benchmarked using available properties (i.e., density, bond lengths, angles, diffusivity, and viscosity).

Nuclear Criticality Safety Program (NCSP)↗

Results for Gamma-Ray Spectroscopy Characterization of Residual Radioactivity on Personal Protective Clothing

In early 2019, Health Physics Services (HPS) personnel put together a test plan to use the shielded gamma-ray spectroscopy instruments in the Health Physics Services Technical Laboratory (HPSTL) to perform a series of measurements on laundered personal protective clothing used in radiological areas at the Savannah River Site (SRS). That test plan describes the methodology used to determine the quantities of laundry to be obtained from operational facilities, the methods for quality control checking the instruments throughout the measurement period, and the physical arrangement of samples in the counting shields. The test plan is documented in SRNS-J6700-2019-00037 (Reference 1).

61 RADIATION PROTECTION AND DOSIMETRY↗

Analysis of Second Target Station Target Segment Removal Dose Rates

This report documents the analysis of the dose rate fields in two configurations of the target system of the Second Target Station (STS) at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) in support of the remote handling target segment removal operation. The two configurations detailed in this report are both assuming 8 hours since beam was incident on the target. The first configuration assumes all shielding is in place just as it was during operation and allows for confirmation of hands-on maintenance capability of the target drive components. The second configuration assumes that a hatch in the target drive room ceiling has been opened to the high bay, the target segment removable core vessel shield block has been removed, and there is a direct line-of-sight to the target segment some 5 m below the target drive room elevation. Dose rate fields in both configurations are documented in this report along with details of the methods of computation. The dose rate fields included in this report are intended to support the engineering of remote handling special equipment and the planning of remote handing operations. The models and analysis detailed in this report are in support of the preliminary design of the STS and future analysis is needed to support the final design of the STS.

61 RADIATION PROTECTION AND DOSIMETRY↗

Ultrasound preliminary cyber-physical evaluation

In this assessment, we focused on identifying unwanted RF emissions from the subject devices. This consisted of two major approaches: 1) Monitor for emitted RF during operation and ensure it is within expected bounds. 2) Inspect circuit board construction for any geometry or design patterns that could lead to RF emissions, whether intentional or unintentional. These patterns include: a) Single-ended PCB traces that could operate as an antenna. b) Insufficient shielding around typically “noisy” components such as switching power supplies. c) Free-hanging high-speed signal wires with no shielding that could emit RF. Suspicious components and design patterns were given a plausible reason to be included in the design. Any suspicious components or patterns warranted reason for deeper investigation. After more in-depth analysis no components were found to be intentionally malicious or had unexpected functionality.

42 ENGINEERING↗

Capability for time-resolved diffraction for material phase in extreme conditions (Q2 FY24 Report)

After a series of development experiments, a new x-ray diffraction diagnostic called FIDDLE (Flexible Imaging Diffraction Diagnostic for Laser Experiments) has been designed, built, and fielded on four shots at NIF. Here we document the initial performance of FIDDLE fielded as the detector in the time-resolved x-ray diffraction experimental platform. Early challenges with EMP have been documented using offline testing in the Target Diagnostics Factory to identify EMP leaks. These techniques were used to guide the application of shielding to weak spots where EM radiation was able to penetrate inside the diagnostic. Application of shielding was found to be effective at reducing EMP leaks, enabling detection of x-ray diffraction on three temporally separated frames during a high pressure (~1Mbar) experiment where lead (Pb) was observed to undergo a phase transition from HCP to BCC.

36 MATERIALS SCIENCE↗

Estimate of Gamma Dose Rates from Arrays of Fermi-1 Blanket Elements During the MEDE Process

The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

FY21 Progress Report for Advanced Re fabrication/Re-instrumentation Capability Development

In support of performing follow on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and reinstrumentation capability. As a result, the United States Department of Energy has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory. The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of re-instrumentation and the progress to date. Halden spent nearly 30 years developing both refabrication and reinstrumentation. Collaboration with Halden is allowing INL to develop this capability much more rapidly. The results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Microfabricated Ion Traps on Sapphire for Larger Trap Areas and Higher Qubit Count

Surface ion traps are a promising platform for quantum computing due to their potential to store large numbers of ions that can be addressed by electrical and optical control signals in order to implement quantum algorithms. Increasing the power of the quantum computer requires increasing the number of ions, but this poses a significant challenge in that it leads to a non-linear increase in on-chip power dissipation. The primary contributor to this power scaling in current devices is the capacitance between the radio frequency (RF) electrode and the metal plane that shields the silicon substrate from the RF signals applied to it. Silicon has traditionally been chosen for the substrate material for compatibility with the processing required for multi-metal-level traps. In this work, we address these capacitance and fabrication challenges by replacing the commonly used silicon substrate with an insulating sapphire substrate to fabricate a multi-metal-level ion trap, while still employing common semiconductor manufacturing techniques. This change in substrate allows the design to remove the metal shielding from the device design, reducing the capacitance of the RF electrode. The electrical characteristics of these traps were measured, specifically trap impedance, capacitance, and voltage breakdown, and compared to nearly identical silicon trap devices. Finally, we used laser cutting techniques to shape a sapphire wafer into bowtie shapes matching silicon traps previously fabricated at Sandia National Labs to explore solutions for integrating sapphire substrates into non-rectangular ion trap designs.

97 MATHEMATICS AND COMPUTING↗

Heat load measurements for the PIP-II pHB650 cryomodule

This study presents a brief overview of the 1st and 2nd phases and an in-depth analysis of the 3rd phase heat load testing performed on the pHB650 (prototype High Beta 650 MHz) cryomodule at PIP2IT (PIP-II Injector Test Facility), with a focus on both the results and the methodological advancements that have improved testing efficiency and accuracy. A key challenge identified in the testing campaign is the higher-than-expected heat loads observed in the first PIP-II (Proton Improvement Plan II) prototype cryomodules (pSSR1 and pHB650) tested at PIP2IT. Elevated heat loads are concerning given the fixed capacity of the PIP-II cryoplant that is currently being installed at Fermilab. However, understanding the sources of these elevated heat loads offers a critical opportunity to implement effective heat load mitigations on upcoming PIP-II cryomodules to stay within the available capacity of the PIP-II cryoplant. The study includes a summary of test results, descriptions of measurement procedures, and key observations on parameters directly and indirectly related to heat load measurements. Direct observations include measured heat loads and the effectiveness of JT heat exchanger under varying conditions, while indirect observation analyze factors such as the temperature distribution on the two-phase pipe and relief piping under varying conditions. Thermal acoustic oscillations (TAO) were identified during testing, which was mitigated by replacing the original G10 stem with a stainless steel stem equipped with wipers for the cryomodule cooldown valve. A major innovation during pHB650 Phase 3 testing was the development of an automated Python script to streamline data acquisition, analysis, and reporting of heat load results. This script automatically retrieved data from ACNET (Accelerator Control Network), performed heat load calculations, and generated detailed reports featuring plots and tables. This advancement significantly reduced manual labor and enhanced the thoroughness of data analysis compared to earlier campaigns. The heat load test reports were promptly uploaded to the electronic logbook shortly after each test, enabling rapid feedback and collaboration between the SRF and cryogenic teams. The heat load measurements included various components: HTTS (high-temperature thermal shield), LTTS (low-temperature thermal shield), 2K isothermal and non-isothermal heat loads. Results were recorded both within the cryomodule and between the bayonet can supply and return. Measurements were conducted under different operating conditions such as "standard", "linac", and "simulated dynamic". Additionally, HTTS and LTTS heat loads were calculated in real time, allowing for the tracking of thermal stability and identification of changes during testing, both in steady-state and transient conditions. The results of this testing campaign not only provide valuable insights into the performance of the pHB650 cryomodule but also highlight best practices and lessons learned that will inform future cryomodule testing at PIP2IT. These include adopting automated tools for data analysis, refining real-time measurement capabilities, and emphasizing detailed pre-test planning. The framework established in this campaign aims to set an improved standard for cryomodule testing and heat load reporting in future cryomodule test campaigns.

Porwisiak, D. [Fermilab; Wroclaw Tech. U.]↗

An Advanced Machine Learning and Artificial Intelligence System for Demonstrating Radiation Regulatory Compliance in DOE Accelerator Facilities

In this Phase II proposal, Applied Research LLC (ARLLC), Thomas Jefferson National Accelerator Facility (Jefferson Lab), and Old Dominion University (ODU) propose the combination of domain knowledge (beam characteristics, fixed structural shielding, earthen burden (the soil and foliage added to the dome of the experimental halls as additional shielding), etc.), machine learning (ML) and/or artificial intelligence (AI) to correlate a variety of multi-modal onsite signals and the radiation fields seen in accessible areas of the accelerator site and the site boundary. The ML/AI will consider the complex influence of environmental parameters affecting the radon contribution of the measurements, focusing on actual data obtained from Jefferson Lab. In Phase I, the coded beam and location data were fed into a deep learning model to predict doses at several designated locations in Jefferson Lab’s facility. Moreover, a dense radiation map was generated using only a sparse collection of the samples in a facility. In Phase II, we will develop a software prototype containing a radiation prediction algorithm, dense radiation map algorithms, and background noise prediction algorithms, with actual data used to evaluate the prototype. This work will provide a framework for evaluation of radiation measurement results around the site based on learned responses. In addition, the proposed approach allows more granular mapping of radiation levels. Better understanding and communication of these levels is related to the overall approach in keeping doses to personnel ALARA.

43 PARTICLE ACCELERATORS↗

Evaluation of Second Target Station Target Streaming for the Preliminary Design of the Target Assembly

This report documents the analysis of dose rate fields from a combination of two sources stemming from the target system of the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source (SNS) Second Target Station (STS). This analysis supports the target drive components’ lifetime estimates and conventional facilities’ preliminary design. The two sources detailed in this report are the radiation generated by the proton beam incident on the target and the activated cooling water in the pipes at the top of the target assembly. The analysis considers the STS target system in operational configuration, with all shielding in place. The dose rate fields are calculated in the target drive room and through the target drive room ceiling. This report calculates the thickness necessary for the target drive room’s high-density concrete ceiling to meet the 0.25 mrem/hr dose limit requirement in the high bay directly above the target drive room. In addition to sizing the target drive room ceiling, this report includes details of the dose field surrounding the target drive assembly to estimate the deposited dose in the target drive assembly components resulting from both streaming and decay of activated water. Prompt neutrons and gammas generated by the proton beam incident on the target primarily stream along the gaps between the target assembly and core vessel shielding. Because the models and analysis detailed in this report are based on the STS preliminary design, future analysis is needed to support the STS final design.

43 PARTICLE ACCELERATORS↗

LBNF Material Activation Study

This poster presents a study on material activation at the Fermilab Long-Baseline Neutrino Facility (LBNF), focusing on understanding how high-energy beams interact with surrounding materials to produce radioactive isotopes. Using a simplified model of the LBNF-20 bunker and the FLUKA Monte Carlo simulation tool, the project quantifies isotope production at key locations inside and outside the shielding structures. The study employs a two-step simulation process, tracking primary and secondary particles, and explores the impact of different geometric and material configurations on activation rates. Results include preliminary comparisons of simulated activation rates and suggest methods for refining geometry, improving accuracy, and efficiently simulating future configurations. This work informs the design and operation of the LBNF facility, aiding in radiation safety and shielding optimization. Future efforts will focus on sample iteration, automated simulations, and detailed residual isotope analyses to further enhance the understanding of material activation in high-energy physics environments.

Pietraszek, Jacob [Unlisted, US]↗

Foam and Unreacted Material Reduction in Magnesium Oxysulfate for SRPPF Waste Solidification

A magnesium oxysulfate (MOS) grout formulation has been proposed to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). This formulation uses a combination of light-burnt MgO, anhydrous MgSO 4 , and dead-burnt MgO resulting in a grout with good mixability and acceptable density. The setting time for the grout is within the operational limits of the SRPPF facility and the leachate pH is ~9.4, which is within the assumed pH range of the brine from the Waste Isolation Pilot Plant (WIPP). When first tested at the 1-gallon small-scale and 55-gallon full-scale, there was a significant foam layer that raised concerns of a nonhomogeneous final form. A nonhomogeneous mixture could have variable density throughout the waste form and therefore allow for unequal shielding. To reduce the potential for unequal shielding in the final form, a series of foam reduction tests were carried out and a final formulation of 28 wt% light burnt MgO, 10 wt% MgSO 4 , and 62 wt% dead burnt MgO was used to remove the foam in under two hours while still maintaining a mix that has an acceptable density, mix time, peak temperature, and set time.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Improvement and Verification of Online Cross Section Generation Capability of Griffin for TRISO-fueled Reactors

Griffin, a MOOSE-based reactor multiphysics code jointly developed by Idaho National Laboratory and Argonne National Laboratory under the DOE Office of Nuclear Energy’s NEAMS program, has pursued the development of an online multigroup cross section generation capability for a few years to enable high-fidelity, problem-dependent neutronics analyses of advanced thermal reactors. Recent advancements in Griffin’s online multigroup cross section generation capability have significantly improved the accuracy, robustness, and efficiency of self-shielding calculations for both prismatic and pebble-bed TRISO-fueled reactor applications. Key developments include a unified fuel self-shielding method applicable to both TRISO and annular compact/spherical shell fuel zone geometries; an advanced Dancoff Category-based Equivalence Theory using a bell function for non-fuel resonance treatment, achieving more than an order-of-magnitude speedup compared to the Tone method; an on-the-fly multigroup equivalence approach to mitigate group condensation errors; and a streaming correction method for pebble-bed homogenization. A proof-of-concept demonstration of on-the-fly group condensation with consistent P0 transport correction was also achieved. The method reproduced direct fine-group solutions with excellent accuracy (eigenvalue errors within 10 pcm and pin-power differences within 0.5%), but due to performance limitations of the current fixed-source solver, improvements to solver efficiency will be addressed in future work. Verification tests were performed on graphite-moderated TRISO-fueled two-dimensional core benchmark problems representing gas-cooled microreactors, heat pipe-cooled microreactors, gas-cooled pebble-bed reactors, and fluoride salt-cooled high-temperature reactors. Across all cases, Griffin showed excellent agreement with Serpent2 continuous energy Monte Carlo solutions: eigenvalue errors within 200 pcm, pin-power root-mean-square errors within 2%, and control rod and drum worth errors less than 2%. It should be noted that, for the benchmark problem, cross section generation contributed less than 3% of the total simulation times. These results demonstrate that Griffin’s online cross section generation capability delivers accurate and efficient reactor physics solutions across a wide spectrum of TRISO-fueled advanced reactor designs. With further improvements to the fine-group fixed-source solver and planned extensions to depletion, transients, and coupled neutron–gamma transport, Griffin will be well-positioned to become a powerful and comprehensive tool for advanced reactor analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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↗

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4. The experimental data used as basis for validation herein consist of measurement data for nuclide inventories and decay heat, including the following: 1. radiochemical assay (RCA) measurements of nuclides important to burnup credit, decay heat, and radiation shielding in 205 light-water reactor (LWR) spent nuclear fuel samples that cover burnups ranging up to 80 GWd/MTU and initial fuel enrichments up to 4.9% 235U; 2. full-assembly decay heat measurements for 236 LWR assemblies with initial fuel enrichments ranging up to 4% 235 U, assembly burnups of 5–51 GWd/MTU, and decay times after fuel discharge in the 2- to 27-year range (of importance to spent nuclear fuel storage, transportation, and disposal); and 3. pulse fission irradiations for fissionable materials at cooling times of interest to severe accident analyses (<10 5 s).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Design and Development of the Handling Frame Assembly Table for the NIF Sustainment Project (Draft)

The Handling Frame Assembly Table is a custom designed piece of equipment developed to support the NIF Sustainment Project. The NIF Sustainment project's goal is to refurbish the National Ignition Facility (NIF), the world’s largest and most energetic laser, so it can continue advancing research in fusion energy, national security, and high energy density physics. A key aspect of this effort involves replacing all 1,728 blast shields on the NIF, which requires specialized equipment such as the Handling Frame Assembly Table to support the production of new blast shields. This project follows a systems engineering approach that guides the design sequence. This report documents the design process from initial concept through final design, including stakeholder analysis, requirements development, conceptual design, design analysis and validation, and final design. Future work will focus on building and commissioning the system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Quantifying Trapped Magnetic Vortex Losses in Niobium Resonators at mK Temperatures

Trapped magnetic vortices in niobium introduce microwave losses that degrade the performance of superconducting resonators. While such losses have been extensively studied above 1 K, we report here their direct quantification in the millikelvin and low-photon regime relevant to quantum devices. Using a high-quality factor 3D niobium cavity cooled through its superconducting transition in controlled magnetic fields, we isolate vortex-induced losses and find the resistive component of the sensitivity to trapped flux S to be approximately 2 nOhm/mG at 10 mK and 6 GHz. The decay rate is initially dominated by two-level system (TLS) losses from the native niobium pentoxide, with vortex-induced degradation of T1 occurring above Btrap~50 mG. In the absence of the oxide, even 10 mG of trapped flux limits performance, Q0 ~ 1010, or T1 ~350 ms, underscoring the need for stringent magnetic shielding. The resistive sensitivity, S, decreases with temperature and remains largely field-independent, whereas the reactive component, S , exhibits a maximum near 0.8 K. These behaviors are well modeled within the Coffey Clem framework in the zero-creep limit, under the assumption that vortex pinning is enhanced by thermally activated processes. Our results suggest that niobium-based transmon qubits can tolerate vortex-induced dissipation at trapped field levels up to several hundred mG, but achieving long coherence times still requires careful magnetic shielding to suppress lower-field losses from other mechanisms

Bafia, Daniel P. [Fermilab]↗