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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
NASA NTRS1972

The observation of damage regions produced by neutron irradiation in lithium-doped silicon solar cells.

Study regions of lattice disorder produced in lithium-doped float-zone melted n/p-type silicon solar cells by irradiation with monoenergetic neutrons at doses between 10 to the 10th and 10 to the 13th per cu cm. The defect regions were revealed by chemically etching the surface of the solar cells and by observing carbon replicas in an electron microscope. It was found that the defect density increased with increasing irradiation dose and increased lithium content, whereas the average defect diameter was found to decrease. From thermal annealing experiments it was found that in the lithium-doped material the defect structure was stable at temperatures between 300 and 1200 K. This was found to be in contrast to the undoped material where at the lowest doses considerable annealing was observed to occur. These results are discussed in terms of the theoretical predictions and models of defect clusters proposed by Gossick (1959) and Crawford and Cleland (1959).

Ghosh, S.↗
DOE OSTI2023

Characterization of the Fast-Neutron Irradiator and the Fast-Flux Tube Irradiation Fixtures at the Pennsylvania State Breazeale Reactor

Accurate knowledge of the neutron spectrum at a nuclear research reactor is a prerequisite for planning irradiation experiments, as well as for evaluating irradiation exposure results. The neutron-flux spectrum in the fast-neutron irradiator (FNI) and the fast-flux tube (FFT) irradiation fixtures at the Pennsylvania State Breazeale Reactor (PSBR) were characterized using the multi-foil neutron activation method. These irradiation fixtures make use of graded shielding to produce unique neutron fields. Multiple foil sets were irradiated in the fixtures with different exposure times and reactor powers to understand the stability over a wide range of operating conditions. Measured results were evaluated against a MCNP6 simulation to produce a measurement-informed neutron flux-energy spectrum for each fixture using STAYSL_PNNL. Simulated estimates of the FNI fixture, a newer fixture (~25 years old), demonstrated excellent agreement with measured results; the FFT did not. Thermal neutron measurements from the FFT suggest there is additional thermal leakage not captured in the simulation model. In conclusion, possible explanations for the discrepancy include burn-out or degradation (i.e., micro-cracking, gaps, etc.) in the boral and cadmium liners over the lifetime of the fixture (~40 years old).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1968

Resistivity measurements of neutron-irradiated pure metals and Al-Zn alloys

Report presents resistivity measurements and their interpretation for neutron-irradiated pure metals and Al-Zn alloys. The influence of temperature, the role of point defects, and the aging behavior on resistivity are considered. The experimental procedures and results are discussed in detail.

Horak, J. A.↗
DOE OSTI2022

Mechanisms of stored energy release in silicon carbide materials neutron-irradiated at elevated temperatures

Understanding the stored energy release behavior of SiC is crucial for the design of SiC-based fuel cladding for light water reactors because stored energy release affects response during accident conditions. Differential scanning calorimetry was used to evaluate the stored energy of monolithic SiC and SiC fiber–reinforced SiC matrix composite variants following neutron irradiation under light water reactor–relevant dose and temperature conditions. The tests were performed at heating rates of 0.5, 3, and 20 K/min up to 1,273 K. The onset temperature of the energy release roughly corresponded to the irradiation temperature, and energy release was delayed as the heating rate increased. The monolithic and composite specimens exhibited similar stored energy release behavior up to ~1,000 K, and above that temperature, the energy release depended more on material type. The energy release measured in this study is below the material-specific heat capacity at temperatures below ~1,000 K but exceeds it under certain material and annealing conditions at temperatures above ~1,000 K. The amount of irradiation-induced volumetric swelling of the specimens was an indication of the total stored energy. Based on this observation, the energy release behavior was described by a swelling recovery model.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Rare-gas analyses on neutron irradiated Apollo 12 samples.

Argon, krypton, and xenon from stepwise heating of five Apollo 12 rocks that had been irradiated to a neutron fluence of about 10 to the 19th neutrons/sq cm were analyzed mass-spectrometrically. The Ar40-Ar39 ages were determined and range from 3.18 plus or minus 0.06 x 10 to the 9th years to 3.32 plus or minus 0.06 x 10 to the 9th years. Trace elements Ba, Br, I, and U were measured via the neutron-induced reactions that produce isotopes of krypton and xenon. The following concentration ranges were observed: Ba, 46 to 70 ppm; Br, 48 to 146 ppb; I, 16 to 73 ppb; and U, 170 to 247 ppb.

Alexander, E. C., Jr.↗
DOE OSTI2026

Alternative Method for Performing As-Built, As-Irradiated Neutronics Analyses

Neutronics analyses have been used to assist in the design, development, and qualification of miniature fuel (MiniFuel) experiments irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) since their conception almost a decade ago. In recent years, upon completion of a MiniFuel target’s irradiation, neutronics analyses that reflect the as-built, as-irradiated conditions of an experiment have also been performed to validate postirradiation examination (PIE) efforts. Several radiation transport and depletion/activation software wrapper programs have been developed to increase the fidelity of these calculations; at ORNL, a program called HFIRCON is the primary software method employed. However, the utilization of these wrapper programs comes at a cost. To produce accurate results, the radiation transport portion of these analyses requires significant time and resources to execute. Therefore, this report discusses an alternative methodology for calculating some of the desired output for PIE validation—specifically, the output used in evaluating the experiment’s fuel burnup—without complete reliance on ORNL’s wrapper program. The alternative method is not a new software but rather a procedure for utilizing the depletion/activation software, SCALE, in conjunction with limited input data and reasonable assumptions to bypass most of the radiation transport calculations and thereby reduce the time required for the completion of a standard analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

The Effects of Manufacturing Techniques on Neutron-Irradiated C26M Mechanical Properties

Nuclear energy provides 13% of the total energy produced globally. To further improve the safety and reliability of nuclear reactors material challenges need to be successfully resolved. C26M/FeCrAl alloy systems can improve the safety and reliability of current Generation III reactors due to the potential benefits of a stable passivation layer from the Al, which can protect against corrosion, and promising mechanical performance at boiling- and light-water reactor temperatures. However, information is limited on how manufacturing techniques affect the mechanical performance of C26M in neutron-irradiated environments. In this research, C26M specimens were manufactured using three techniques and then irradiated at the Advanced Test Reactor at Idaho National Laboratory. This research aims to bridge this gap by investigating how the manufacturing techniques affects the mechanical performance of irradiated C26M at Generation III reactor relevant temperatures.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

Infrared MBE-Grown HgCdTe Focal Plane Arrays and Cameras After High Energy Neutron Irradiation

HgCdTe is one of the most important materials for the fabrication of infrared detectors and focal plane arrays (FPAs) deployed in environments where high-energy particles, such as protons and neutrons, are present. We designed and fabricated HgCdTe-based FPAs that can be used in high neutron radiation environments and we measured their characteristics. The influence of the radiation on the infrared FPAs and cameras is investigated. HgCdTe material and devices are capable of maintaining high performances in a high energy neutron irradiation environment. For MWIR FPA directly facing a $2.59 × 10^8$ n/cm$^2$ s neutron flux beam (with the highest energy 66 MeV) for 1 h, the noise equivalent differential temperature (NEDT) increased ~8 times after irradiation. However, NEDT decreased to 33 mK (compared to the original value of 21 mK) after one warming-up (to room temperature) and cooling-down cycle. The NEDT for the MWIR FPAs mounted parallel to the beam did not degrade (16 mK and 28 mK before irradiation, changed to 18 mK and 26 mK after irradiation, respectively).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Comprehensive Characterization of Helium-Induced Degradation of the Friction Stir Weld on Neutron-Irradiated 304L Stainless Steel

The report describes new experimental results on the mechanical performance of the friction stir welds made on neutron-irradiated 304L stainless steel with helium. The report focuses on helium-related issues (e.g., the helium-induced degradation in the welded joint), aiming at the need to repair irradiated components of nuclear power plants. The friction stir welds analyzed here were previously produced at the US Department of Energy’s Oak Ridge National Laboratory, and initial characterization work was performed, mostly addressing the microstructure and if macroscopic cracks and helium bubbles present. The present work attempts to perform a more comprehensive study to assess mechanical performance (i.e., microhardness distribution, tensile properties, tensile deformation behaviors, and fractography analysis).

36 MATERIALS SCIENCE↗
NASA NTRS1972

Observation of defect annealing in neutron-irradiated silicon by space-charge-limited current.

Use of space-charge-limited current (SCLC) to observe defect annealing in neutron-irradiated silicon. From the observation of the trap-filled-limit voltage a deep-trap concentration in the irradiated unannealed crystals is found to be at least 1.4 x 10 to the 12th power per cu cm. Shallow traps are also observed. This deep-trap concentration decreases with anneal temperature up to 200 C, above which SCLC can no longer be observed. There is evidence that at least some of this decrease is due to the conversion of the deep traps to shallow traps.

Magee, T. J.↗
NASA NTRS1991

Segmented Ge detector rejection of internal beta activity produced by neutron irradiation

Future Ge spectrometers flown in space to observe cosmic gamma-ray sources will incorporate segmented detectors to reduce the background from radioactivity produced by energetic particle reactions. To demonstrate the effectiveness of a segmented Ge detector in rejecting background events due to the beta decay of internal radioactivity, a laboratory experiment has been carried out in which radioactivity was produced in the detector by neutron irradiation. A Cf-252 source of neutrons was used to produce, by neutron capture on Ge-74 (36.5 percent of natural Ge) in the detector itself, Ge-75 (t sub 1/2 = 82.78 min), which decays by beta emission with a maximum electron kinetic energy of 1188 keV. By requiring that an ionizing event deposit energy in two or more of the five segments of the detector, each about 1-cm thick, the beta particles, which have a range of about 1-mm, are rejected, while most external gamma rays incident on the detector are counted. Analysis of this experiment indicates that over 85 percent of the beta events from the decay of Ge-75 are rejected, which is in good agreement with Monte Carlo calculations.

Varnell, L. S.↗
DOE OSTI2026

Site specific porosity-thermal performance correlations in neutron irradiated U-10Zr fuel

In this work, we present a site-specific three-dimensional analysis of porosity evolution in neutron-irradiated U–10Zr metallic fuel using high-resolution synchrotron X-ray tomography. Focused ion beam cubic lift-outs from four radial positions—fuel center, middle, edge, and fuel-cladding interaction (FCCI) zones—were imaged, reconstructed, and segmented to quantify pore volume, density, morphology, and pore connectivity. Porosity increased modestly from 5.5% at the center to 10.5% at the edge, yet pore number density increased by over two orders of magnitude in the fuel portion near the FCCI interface (from 4.8 x 10 4 to 6.0 x 10 6 pores/mm 3 ). Morphological classification revealed a progression from small spherical pores at the center to equiaxed and tortuous networks at the periphery, with enhanced orientation along the radial direction. Connectivity and permeability analysis reveal that the FCCI region maintains dense, highly interconnected pores despite a moderate volume fraction, enabling rapid fission gas transport and lanthanide migration. Effective thermal conductivity models incorporating sodium logging confirm that these site-specific pore features critically influenced heat transport during reactor irradiation. These findings demonstrate that pore topology—not just porosity fraction—controls thermal performance and thermal pathways in U–10Zr fuels, impacting fuel thermal performance in a reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Computational Optimization of 133m Xe Production via Neutron Irradiation in a TRIGA Reactor

Here, the Comprehensive Nuclear-Test-Ban Treaty bans all nuclear tests worldwide. As part of treaty compliance, the concentration of radioactive nuclides in the atmosphere is monitored to detect nuclear weapons tests. Radioactive noble gas fission products, specifically radioxenon, can vent into the atmosphere after a nuclear weapons test, even if the test is well contained underground or underwater. Radioxenon thus serves as a signal for nuclear weapons tests. All atmospheric monitoring systems require samples of radioxenon isotopes for detector calibration, quality control, and certification. Here, we present a novel, improved method for creating samples of 133m Xe via neutron irradiation of 132 Xe in the Washington State University TRIGA reactor. 132 Xe neutron absorption results in either 133 Xe or 133m Xe—thermal neutron absorption results in 133m Xe 12% of the time, while fast neutron absorption (above ~1 MeV) results in 133m Xe ~50% of the time. To optimize the production of 133m Xe via neutron absorption in 132 Xe in the thermal TRIGA reactor, spectral tuning using an irradiation chamber is required to maximize the fraction of fast neutrons being absorbed and minimize the number of thermal neutrons interacting with the 132 Xe. We used MCNP to tally 132 Xe absorptions with the isotopic tally function, flux tallies and neutron attenuation to estimate the number of neutrons reaching the 132 Xe through the irradiation chamber, and the adjoint importance function to improve the source strength estimate. Additionally, we performed a heat transfer analysis for safety considerations. It was determined that the use of a 96% enriched 10 B boron carbide chamber, placed next to the fuel elements in reactor position D8, increases the 133m Xe/ 133 Xe activity ratio from a baseline value of 0.3 to 1.0, a 233% increase. Additionally, it was determined that the alpha heating produced in the boron does not become an unmanageable problem in the Washington State University reactor.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2021

Microstructure and Mechanical Performance of the Friction Stir Welds Made on Neutron-Irradiated Steel with Helium

This report describes new experimental results on the microstructure and mechanical performance of the friction stir welds made on neutron-irradiated steel with helium. The report focuses on helium-related issues, specifically, helium-induced degradation in the welded joint, aiming to repair irradiated components of nuclear power plants (NPPs). The friction stir welds analyzed here were produced at Oak Ridge National Laboratory previously, and limited characterization work was performed, mostly addressing the presence or absence of macroscopic cracks. The present work attempts to perform a more comprehensive study to assess microstructure conditions, grain size, plastic strain gradients, the morphology of the helium-induced damage, and deformation behavior.

42 ENGINEERING↗
DOE OSTI2020

Fullerene-like defects in high-temperature neutron-irradiated nuclear graphite

Irradiation-induced defect evolution in graphite is particularly important for its application in graphite-moderated nuclear reactors. The evolution of defects directly influences macroscopically observed property changes in irradiated nuclear graphite which, in turn, can govern the lifetime of graphite components. This article reports novel defect structures and the irradiation response of microstructural features occurring in high-temperature irradiated nuclear graphite IG-110. High resolution transmission electron microscopy (HRTEM) was used to characterize specimens neutron-irradiated at a high temperature (≥ 800°C) at doses of 1.73 and 3.56 atomic displacements per atom (dpa). Concentric shelled and fullerene-like defects were found to result in swelling along the c-axis and contraction along the a/b-axis of crystallites. Furthermore, such defects are shown to occur within, and partially fill, Mrozowski cracks prior to turnaround dose. In addition, in situ TEM under similar irradiation conditions was used to capture the real-time dynamic evolution of defects, providing unambiguous analysis of the evolution of the graphite structures during irradiation. Here, the results suggest the mainstream theory for radiation damage in nuclear graphite (which assumes additional basal plane formation as the sole reason) to be an incorrect interpretation of defect evolution contributing to irradiation-induced property changes at higher temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1995

Effect of neutron irradiation on the London penetration depth for polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor

Magnetization studies of polycrystalline Bi(1.8)Pb(0.3)Sr2Ca2Cu3O10 superconductor, prior to and after neutron irradiation, showed an increase in J(sub c) due to irradiation damage. Analysis of the equilibrium magnetization revealed significant increases in other more fundamental properties. In particular, the London penetration depth increased by approximately 15 percent following irradiation with 8 x 10(exp 16) neutrons/sq cm. Corresponding changes were observed in the upper critical magnetic field H(sub c2). However, the most fundamental thermodynamic property, the superconductive condensation energy F(sub c), was unaffected by the moderate level of neutron-induced damage.

Ossandon, J. G.↗