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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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Leveraging Afterglow in Scintillation-based x-ray detectors for spacetime-resolved computed tomography for accelerated acquisition and high-speed event capture

Afterglow in x-ray imaging for high-speed radiography is a constraint that limits imaging systems to low-light/fast decay screens which create poor data. Current approaches focus purely on using low-light yield screens with fast decay to avoid multiple exposure pileup due to afterglow. The goal of this work is to develop a statistical estimation approach to leverage afterglow to improve image quality thus allowing for higher quality imaging components to be used. This will allow for bright screens will slow decay to be used, and then a post-processing step applies the statistical estimation to separate each frame with superior signal compared to low-light/fast decay screens.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Σ3(111) Grain Boundaries Accelerate Hydrogen Insertion into Palladium Nanostructures

Grain boundaries (GBs) are frequently implicated as key defect structures facilitating metal hydride formation, yet their specific role remains poorly understood due to their structural complexity. Here, we investigate hydrogen insertion in Pd nanostructures enriched with well-defined Σ3(111) GBs (Pd GB ) synthesized via electrolysis-driven nanoparticle assembly. In situ synchrotron X-ray diffraction reveals that Pd GB exhibits dramatically accelerated hydriding and dehydriding kinetics compared with ligand-free and ligand-capped Pd nanoparticles with similar crystallite sizes. Strain mapping using environmental transmission electron microscopy shows that strain is highly localized at GBs and intensifies upon hydrogen exposure, indicating preferential hydrogen insertion along GB sites. Density functional theory calculations provide mechanistic insight supporting these findings, showing that hydrogen insertion near Σ3(111) GBs is energetically more favorable and that tensile strain lowers insertion barriers. Furthermore, these results provide atomic-level insights into the role of GBs in hydride formation and suggest new design strategies for GB-engineered Pd-based functional materials.

Anions↗
DOE OSTI2025

First Joint Oscillation Analysis of Super-Kamiokande Atmospheric and T2K Accelerator Neutrino Data

The Super-Kamiokande and T2K Collaborations present a joint measurement of neutrino oscillation parameters from their atmospheric and beam neutrino data. It uses a common interaction model for events overlapping in neutrino energy and correlated detector systematic uncertainties between the two datasets, which are found to be compatible. Using 3244.4 days of atmospheric data and a beam exposure of 19.7⁢(16.3) × 10 20 protons on target in (anti)neutrino mode, the analysis finds a 1.9⁢𝜎 exclusion of 𝐶⁢𝑃 conservation (defined as 𝐽 𝐶⁢𝑃 = 0) and a 1.2⁢𝜎 exclusion of the inverted mass ordering.

Neutrino detection↗
DOE OSTI2023

Novel pyrolytic boron nitride coating to reduce graphite interactions in molten fluoride salt

Future salt reactors like the Fluoride-salt-cooled-High Temperature Reactor require varying material flavors. Graphite is required for neutron moderation and potential fuel forms, while structural metals are required for containment. Studies show graphite/metal interactions exist when both materials share a salt medium. These interactions may accelerate the corrosion of the metal and change the surface of graphite materials. A novel pyrolytic boron nitride coating on graphite is investigated for reducing graphite/metal interactions. Finally, this is the first known application of such a coating, and salt exposure tests reveal the pyrolytic boron nitride coating reduces the interactions between stainless steel and uncoated graphite.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Biochemical responses of hairgrass (Deschampsia caespitosa) to hydrological change

Plant growth and development are closely related to water availability. Water deficit and water excess are detrimental to plants, causing a series of damage to plant morphology, physiological and biochemical processes. In the long evolutionary process, plants have evolved an array of complex mechanisms to combat against stressful conditions. In the present study, the duration-dependent changes in ascorbate (AsA) and glutathione (GSH) contents and activities of enzymes involved in the AsA-GSH cycle in hairgrass ( Deschampsia caespitosa ) in response to water stress was investigated in a pot trial using a complete random block design. The treatments were as follows: (1) heavily waterlogging, (2) moderate waterlogging, (3) light waterlogging, (4) light drought, (5) moderate drought, (6) heavily drought, and (7) a control (CK) with plant be maintained at optimum water availability. The hairgrass plants were subjected to waterlogging or drought for 7, 14, 21 and 28 days and data were measured following treatment. Results revealed that hairgrass subjected to water stress can stimulate enzymatic activities of ascorbate peroxidase (APX), glutathione peroxidase (GPX), glutathione reductase (GR), dehydroascorbate reductase (DHAR), monodehydroascorbate reductase (MDHAR) and L-galactono-1, 4-lactone dehydrogenase (GalLDH), switched on the ascorbate-glutathione (AsA-GSH) cycle and the L-galactose synthesis, up-regulated the contents of AsA and GSH, and maintained higher ratios of ascorbate to dehydroascorbate (AsA/DHA) and reduced glutathione to oxidized glutathione (GSH/GSSG) to alleviate potential oxidative damage. However, the light waterlogging did not induce hairgrass under stress to switch on the AsA-GSH pathway. In general, the critic substances and enzyme activities in AsA-GSH metabolic pathway increased as the increase of water stress intensity. As the increase of exposure duration, the critic antioxidant substances content and enzyme activities increased first and then maintained a relatively stable higher level. Our findings provide comprehensive information on biochemical responses of hairgrass to hydrological change, which would be a major step for accelerating ecological restoration of degradation alpine marshes in the Qinghai-Tibetan Plateau.

Luo, Qiaoyu↗
DOE OSTI2023

Measurement Of The Total Cross-Section Of Muon Neutrino Charged-Current Coherent Pion Production in NOvA Near Detector

Charged Current (CC) coherent neutrino-nucleus pion production is characterized by small momentum transferred to the nucleus, which is left in its ground state. Despite the relatively large uncertainties on the production cross-section, coherent production of mesons by neutrinos represents an important process, as it can shed light on the structure of the weak current and can also constitute a potential source of background for modern neutrino oscillation experiments and searches for Beyond Standard Model (BSM) physics. This Ph.D. thesis presents a new measurement of CC coherent pion production in the NOvA near detector at the Fermi National Accelerator Laboratory (Fermilab). The analysis is based on the use of both particle identification and kinematic selection criteria based on Convolutional Neural Networks (CNN) to achieve a selection purity of 60%. Given the energy range 1-5 GeV accessible with the available NOvA exposure in the NuMI beam, the results will also be relevant for f uture neutrino experiments like the Deep Underground Neutrino Experiment (DUNE).

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Characterizing the flux effect on the irradiation embrittlement of reactor pressure vessel steels using machine learning

In-service exposure to high-energy neutrons embrittles reactor pressure vessel (RPV) steels. An increase in the yield stress (Δσ y ) results in a corresponding increase in the brittle to ductile transition temperature (ΔT c ). Most existing models underpredict ΔT c at higher fluence following accelerated irradiations in test reactors. High fluence, up to 10 20 n/cm 2 in some cases, will be reached over extended RPV vessel operation of 80 years, or more, at low service flux. Embrittlement has been extensively studied in accelerated, higher flux test reaction irradiations. However, the use of test reactor data naturally raises the question of flux effects. Here, this study used a machine learning approach trained on a set of hardening data, covering a wide range of flux, fluence, and steel compositions to determine the interactive effects of both irradiation and material variables on Δσ y . The analysis included machine learning-based cross-plots of the variable dependence of Δσ y for six core steels (i.e., CM6, LC, LD, LG, LH, and LI), with controlled differences in their Cu and Ni contents. A primary objective is to evaluate an effective to actual fluence (Φt e /Φt) ratio, as a function of flux, fluence, and steel composition. This is information critical to properly use intermediate flux-high fluence data in calibrating a low flux-high fluence embrittlement model. The predicted Φt e /Φt is reasonably consistent with estimates previously derived from a physics-based solute recombination trap model.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Potential induced degradation in c-Si glass-glass modules after extended damp heat stress

Traditional Glass-Backsheet (GB) photovoltaic (PV) modules have been the industry standard for a long time, but the Glass-Glass (GG) modules are quickly rising in popularity. PV modules installed in hot-humid climates with high string voltages can undergo potential induced degradation (PID). So far, to the best of our knowledge, only fresh modules with strong interfacial adhesion have been investigated for PID. However, in reality, the PV modules have weak interfacial adhesion after a few years of field exposure. Therefore, it is essential to evaluate PV modules with weakened interfaces. In this study, we investigated the PID susceptibility of PV modules with weakened interfaces after subjecting them to 2000 hours of damp heat (DH2000) at 85°C/85% relative humidity (RH) in an accelerated environmental chamber. Fresh GG modules were also stressed for PID to compare with PID degradation of DH-stressed modules. Pre- and post-characterization tests were done before, between, and after each stress method to determine the changes in electrical performance, cell metallization properties, and hotspot properties. It is observed that fresh GG modules showed little/no degradation (less than 1%) in maximum power (Pmax), whereas the GG modules that underwent sequential DH and PID degraded by 11% to 12%. Potential mechanisms for these degradations are also presented. Furthermore, the results presented in this study are critical for the industry, considering that the bifacial modules with GG construction will be dominant in the next 10 years.

14 SOLAR ENERGY↗
DOE OSTI2026

Compatibility of Type 304H Stainless Steel in Static Molten FLiBe for Inertial Confinement Fusion Reactors: Role of Impurities and Redox Control

Molten fluoride salts, such as FLiBe (LiF-BeF2), are promising candidates for tritium breeding and heat transfer in fusion reactors, but corrosion of structural materials remains a major challenge. This study investigates the corrosion behavior of austenitic stainless steel 304H in purified and NiF2-containing FLiBe at 500°C and 600°C, focusing on the effects of impurities and redox control. Exposure to purified FLiBe resulted in the concurrent depletion of Cr, Mn, and Fe, with corrosion at 500 °C dominated by the combined oxide formation and elemental dissolution, while at 600°C elemental depletion was predominant. The addition of a controlled NiF2 impurity significantly accelerated corrosion at both temperatures, demonstrating the sensitivity of 304H to the salt redox state. Beryllium additions were effective in mitigating corrosion for both baseline and NiF2-containing FLiBe; minimal depletion of Cr, Mn, and Fe occurred with Be additions as low as 2.5 mg (147 wppm), and no NiBe intermetallics formed at 5 mg (294 wppm), indicating that small Be inventories can provide substantial protection without deleterious phase formation. Thermodynamic equilibrium and coupled thermodynamic-kinetic analyses at the salt-alloy interface suggested low corrosion rates in systems with limited hydrogen fluoride (HF) generation, highlighting the importance of salt redox control. Estimates for a Be addition rate were calculated for the HYLIFE-II fusion reactor that can mitigate corrosion-induced degradation, assuming complete conversion of tritium to tritium fluoride (TF). Overall, 304H shows reasonable compatibility with FLiBe under optimized redox conditions. These results provide quantitative guidance for material selection and salt management in fusion blanket and heat exchanger systems and motivate validation under flowing, nonisothermal, and irradiated conditions.

Pillai, Rishi [ORNL] (ORCID:0000000243688197)↗
DOE OSTI2025

Decoupling Li out-diffusion and surface diffusion in the lithiation-assisted epitaxial growth of lithium tungstate

Lithiation-assisted epitaxy offers a flexible and robust approach for synthesizing high-quality Li-containing materials and interfaces with precise control. Here, in this study, we use lithium tungstate (Li x WO 3+x/2 , where x = 0 to 2) as a model system to investigate the intertwined effects of Li out-diffusion-induced compositional changes and surface-diffusion-induced morphological changes. By systematically varying synthesis and processing conditions, we uncover their impact on lithium tungstate film formation. Comprehensive characterizations, including X-ray diffraction, atomic force microscopy, X-ray photoemission spectroscopy and time-of-flight secondary ion mass spectrometry, reveal that low-temperature growth (< 300 °C) followed by high-temperature annealing yields continuous lithium tungstate films with significantly reduced surface roughness. In contrast, high-temperature deposition (≥ 300 °C) accelerates surface diffusion and Li out-diffusion, leading to island formation. Furthermore, in situ scanning transmission electron microscopy demonstrates the beam sensitivity of Li 2 WO 4 and reveals a phase transition from Li 2 WO 4 to LiWO 3.5 under prolonged electron beam exposure. These findings deepen our understanding of how to control composition and morphology of Li-containing films, providing valuable insights for the design and integration of energy materials.

Shi, Jueli [Pacific Northwest National Laboratory ↗
DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Thermal Aging on Microstructure and Stress Corrosion Cracking Behavior of an Alloy 152 1st Layer Butter Weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40- 80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is critical for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for different durations up to 75,000h (equivalent to 60 years of reactor service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged at 370°C and 450°C to an equivalent of 60 years of service. However, the first weld butter layer has high levels of deformation and is highly susceptible to SCC even in its non-aged condition. Nanohardness testing confirmed the extreme hardening with aging (ΔHV ≅ 100) at two locations within this weldment. In absence of LRO, hardening is suspected to be due to thermally-induced Cr carbide precipitation and coarsening. Testing in a primary water environment of the 1st layer of Alloy 152 weld butter aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. Similarly, the SCC CGR response of the aged weld butter does not appear to show a deterioration in performance, however, the difficult-to-test weldment geometry may affect the test results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Solvari SR, Simplifying Residential Solar to 1 SKU, with Extended Testing

Accelerated weathering and material compatibility testing for polymeric and asphaltic materials. NLR will be testing the durability of the polymeric materials proposed for use in the supporting structures designed by TESCI solar. This will include the bare materials and coupons attached to asphalt shingles. The exposure will be in damp heat and separately heat, humidity and UV light, all followed by mechanical evaluation. Modification 1 is an extension of the first round of testing with both a continuation of some of the same testing and addition of new testing methods and materials. The testing of materials in the condition of A3 looking at mechanical durability will be continued looking at the materials used in the mounting brackets. We will be adding in testing of the silicone adhesive to the module and of the module itself. For the mounting brackets, testing will also be conducted at multiple temperatures and humidity levels to allow for extrapolation to the field.

14 SOLAR ENERGY↗
DOE OSTI2022

Chemical and mechanical interfacial degradation in bifacial glass/glass and glass/transparent backsheet photovoltaic modules

Abstract Glass/glass (G/G) photovoltaic modules are quickly rising in popularity, but the durability of modern G/G packaging has not yet been established. In this work, we examine the interfacial degradation modes in G/G and glass/transparent backsheet modules under damp heat (DH) with and without system bias voltage, comparing emerging polyolefin elastomer (POE) and industry‐standard poly(ethylene‐co‐vinyl acetate) (EVA) encapsulants. We investigate the transport of ionic species at cell/encapsulant interfaces, demonstrating that POE limits both sodium and silver ion migration compared with EVA. Changes to the chemical structures of the encapsulants at the cell/encapsulant interfaces demonstrate that both POE and EVA are more susceptible to degradation in modules with a transparent backsheet than in the G/G configuration. Adhesion testing reveals that POE and EVA have comparable critical debond energies after the DH exposures regardless of system bias polarity. The results of this study indicate that the interfacial degradation mechanisms of G/G appear to be similar to those of conventional glass/backsheet modules. For emerging materials, our results demonstrate that POE offers advantages over EVA but that transparent backsheets may accelerate encapsulant degradation due to increased moisture ingress when compared with the G/G structure.

14 SOLAR ENERGY↗
DOE OSTI2020

Proton irradiation damage and annealing effects in ON Semiconductor J-series silicon photomultipliers

Silicon photomultipliers (SiPMs) have become popular light conversion devices in recent years due to their low bias voltage and sensitivity to wavelengths emitted from common scintillating materials. These properties make them particularly attractive for resource-constrained missions such as space-based detector applications. However, the space radiation environment is known to be particularly harsh on semiconductor devices, where high particle fluences can degrade performance over time. The radiation hardness of a particular SiPM, manufactured by ON Semiconductor (formally SensL), has yet to be studied with high energy protons, which are native to the space radiation environment. As such, to study these effects we have irradiated groups of two SiPMs to four different fluences of 800 MeV protons delivered by the accelerator at the Los Alamos Neutron Science Center. Fluences of 1.68 × 10 9 , 1.73 × 10 10 , 6.91 × 10 10 , and 1.73 × 10 11 protons cm –2 , and their corresponding estimated doses of 0.15, 1.55, 6.19, and 15.5 kRad, were chosen based on estimates of the potential exposure a SiPM might receive during an interplanetary space mission lasting 10 years. We report the effects these doses have on dark current and the self-annealing time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Toward virtual thin-beam 𝜈 cross-section measurements: A feasibility study of 𝜈-Ar interaction analysis with DUNE-PRISM

Neutrino-nucleus cross section measurements are critical for future neutrino oscillation analyses. However, our models to describe them require further refinement, and a deeper understanding of the underlying physics is essential for future neutrino oscillation experiments to realize their ambitious physics goals. Current neutrino cross section measurements provide clear deficiencies in neutrino interaction modeling, but almost all are reported averaged over broad neutrino fluxes, rendering their interpretation challenging. Using the concept of deep underground neutrino experiment-precision reaction independent spectrum measurement (DUNE-PRISM)—a movable near detector that samples multiple off axis positions—neutrino interaction measurements can be used to construct narrow virtual fluxes (less than 100 MeV wide). These fluxes can be used to extract charged-current neutrino-nucleus cross sections as functions of outgoing lepton kinematics within specific neutrino energy ranges. Based on a dedicated simulation with realistic event statistics and flux-related systematic uncertainties, but assuming an almost-perfect detector, we run a feasibility study demonstrating how DUNE-PRISM data can be used to measure muon neutrino charged-current integrated and differential cross sections over narrow fluxes. We find that this approach enables a largely model-independent reconstruction of powerful observables, including energy transfer, typically accessible only in electron scattering measurements, but that large exposures may be required for differential cross section measurements with few-percentage statistical uncertainties.

Neutrino interactions↗