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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Post-Irradiation Evaluation of Eurofer97 Fracture Toughness Using Miniature Multinotch Bend Bar Specimens

In this study, we performed fracture toughness characterization of ten neutron-irradiated Eurofer97 variants using precracked miniature multi-notch bend bar (M4CVN) specimens based on the Master Curve method in the ASTM E1921 standard. The neutron irradiation was performed in the flux trap position of the High Flux Isotope Reactor (HFIR) of the Oak Ridge National Laboratory (ORNL) with the nominal irradiation temperature of 300°C and irradiation dose of 2.5 displacements per atom (dpa). Depending on the irradiation temperature and materials, we observed different degrees of irradiation hardening and embrittlement for ten Eurofer97 variants. The upper shift in the Master Curve reference temperature T0Q vs. the increase in Vickers microhardness values showed a liner relationship for only a few materials indicating different irradiation responses of the Eurofer97 variants.

Chen, Xiang↗
DOE OSTI2020

SNL-NJOY-2016

This is a wrapper around the LANL NJOY-2016 code that interfaces with extended code capabilities and minor modification of the LANL NJOY-2016 code that supports modeling radiation damage to materials. The NJOY Nuclear Data Processing System is a modular computer code designed to read evaluated data in ENDF format, transform the data in various ways, and output the results as libraries designed to be used in various applications. The wrapper provided here permits Sandia-specific control parameters to be used in the input data file. The modifications incorporated here enhance the ability of NJOY to address material damage response functions for many materials, e.g. to include the NRT and arc-dpa forms of the damage energy in addition to the default NJOY-2016 implementation of the sharp-threshold Kinchin-Pease threshold energy treatment. All of the modifications provided here are being made available to the GitHub-based NJOY-2016 code. As useful enhancements found here are incorporated into the baseline NJOY-2016 code, they will be eliminated from this version so as to maintain our compatibility with the baseline NJOY-2016 code. SAND2020-13060 M Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA00035

Griffin, Patrick↗
DOE OSTI2023

IRAD

IRAD (Ion Irradiation and Radiation Damage) is an open-source GUI for SRIM-like vacancy production and implanted ion profile calculations. Using updated databases, IRAD is accessible at https://code.ornl.gov/liny/irad for Windows systems. It performs energy-corrected Iradina calculations, automatically determining dpa and implanted ion concentration, considering total ion fluence. Additional results like ion trajectories, final positions, stopping power distribution, etc., are available. Advanced settings permit customization of parameters such as stopping power database, incident ion angle, and 3D target simulations.

Lin, Yan-Ru [Oak Ridge National Laboratory (ORNL),↗
DOE OSTI2020

Damage Pseudo Cross Section Generator

This software is a python script for generating group damage pseudo-cross-sections for specific elements, and tabulating them in forms useful to python or to office-open workbooks (xlsx). The damage pseudo-cross-section is needed for calculating how much neutron radiation damage a structural material receives in a reactor measured in displacements per atom (DPA). The process of generating these pseudo-cross-sections is laborious as it requires running the nuclear data processing code, NJOY, for hundreds of isotopes, and then post-processing the data. This also allows many opportunities for human error to be involved. This script eliminates these problems by fully automating the process.

Gale, MicahD.↗
DOE OSTI2022

Multiphysics FISPACT-II and TENDL-2019 simulation: neutron-induced damage metrics

Nuclear interactions can be the source of atomic displacement, irradiation-induced defects and transmutation in structural materials. Such quantities are derived from, or can be correlated to, nuclear kinematic simulations of primary atomic energy distributions spectra and the quantification of the numbers of secondary defects produced per primary as a function of the available recoils, residual and emitted, energies. Recoil kinematics of neutral, residual, gas, proton, alpha particle emissions are now more rigorously treated based on recent, complete and enhanced nuclear data parsed in state-of-the-art processing tools. Defect production metrics are the starting point in the complex problem of correlating and simulating the behaviour of materials under irradiation, as experimental information is rare or scattered. Detailed, segregated primary knock-on-atom metrics are now becoming available as the starting point of further simulation processes of isolated and clustered defects in material lattices. This allows more materials, lattices, neutron incident energy ranges, and irradiation conditions to be explored with sufficient data to adequately cover both standard and novel applications and materials: the broader reactor applications landscape. The damage metrics of of materials are systematically explored under typical but different reactor's type environment. The inventory code FISPACT-II combined with the enhanced nuclear data forms of the TENDL-2019 libraries allow one to not only calculate dpa from mostly scattering events but to also properly predict gas production, nuclear heating and transmutation under the same conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Mechanical response of HFIR-irradiated M5FRAMATOME cladding under simple and complex loading conditions

This study investigates the mechanical behavior of High Flux Isotope Reactor (HFIR) irradiated M5FRAMATOME cladding under simple and complex loading conditions through axial tensile and reversible cyclic bending. Axial tension specimens were pre-machined prior to HFIR irradiation while cyclic bend specimens were inserted as intact tubes. Tests articles were neutron-irradiated to 4 and 16 dpa, and specimens were tested at ORNL's hotcell facilities. The axial tension tests were conducted under constant displacement control, and the reversible cyclic bend tests were performed using ORNL's Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) apparatus. Results showed that mechanical response of Cr-coated and uncoated M5FRAMATOME cladding were similar and independent of irradiation dose for axial tension tests, while Cr-coated specimens’ reversible cyclic bend behavior differed from uncoated counterparts. For all tests, irradiation temperature showed a significant impact on the mechanical behavior. Below 280°C, all axial tensile specimens whether coated or not behaved similarly. Above 280°C, YS and UTS showed decrease with increasing irradiation temperature. A similar behavior was also observed in cyclic bend tests as well. The mechanical damage during cyclic bend tests was linked to damage accumulation in unirradiated Cr-coated zircaloy-4 specimens, and the effect of irradiation temperature was related to changing characteristics of defect mobility during high temperature irradiation.

Cinbiz, Nedim [ORNL] (ORCID:0000000346268515)↗
DOE OSTI2025

Irradiation Vehicles for Evaluating SiC/SiC Cladding Bowing Under Neutron Flux Gradients

Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are among the most promising candidates for long term accident-tolerant nuclear fuel cladding. A key challenge related to their deployment is lateral bowing caused by differential radiation-induced swelling under dose or temperature gradients, which could obstruct coolant flow or interfere with control rod/blade movements. Although previous modeling efforts have predicted bowing behavior in light-water reactor (LWR) environments, experimental validation remains limited, especially at prototypic LWR temperatures. This study addresses that gap by irradiating six reduced-length SiC/SiC cladding tubes (~600 mm) in the High Flux Isotope Reactor (HFIR) at ~300°C, which is representative of LWR conditions. The tubes will be housed in a sealed vessel with an inert gas gap to maintain target temperatures and prevent direct coolant contact. Arranged in three pairs, each set will receive a different radiation dose (low, medium, high), with the central pair receiving ~0.1 displacements per atom (dpa)—the expected dose for peak bowing. The experiment will determine the dose-dependent bowing behavior and validate predictive finite element models. In this work, the tubes are freely suspended from pins to allow for unconstrained bowing; however, we present a concept for introducing localized constraints to represent grid spacer effects. Post-irradiation examination will include swelling measurements and profilometry to assess bowing and compare the results with model predictions. This work aims to confirm the conditions under which maximum bowing occurs so as to improve the reliability of SiC/SiC performance models in nuclear applications.

Russell, Nick [ORNL] (ORCID:0000000296099820)↗
DOE OSTI2020

Radiation-Induced Modifications in Copper Oxide Growth

Radiation-induced effects and their influence on oxidation processes were evaluated for their use as a forensic tool for special nuclear material (SNM). A beam of 10 MeV Au³⁺ ions was used to mimic the accumulation of microstructural damage from self-irradiation through the decay of radionuclides. Several copper samples were irradiated as suitable surrogate materials at 200 °C with a flux of 1x10¹² ions/cm·s to damage levels of 5, 10, and 15 displacements per atom (dpa). This corresponds to about 50, 100, and 150 years, respectively, of accumulated α-decay damage in a PuGa alloy assuming a damage rate of 0.1 dpa/year. After irradiation, all samples were exposed to an accelerated aging process induced by thermal treatment at 350 °C for 1 hour in air. This resulted in the growth of a mixed oxide layer (Cu₂O and CuO) which was characterized in detail using several complementary analytical techniques: Scanning Electron Microscopy, Raman spectroscopy, Synchrotron X-ray diffraction (transmission mode), and Grazing Incidence X-ray diffraction. The oxide layer growth of irradiated Cu at 350°C is distinctly modified as a result of the ion irradiation. Most notably, the growth of the CuO phase is suppressed with increasing radiation damage on the Cu substrate, and structural changes occurred in the Cu₂O phase. These results indicate that damage from self-irradiation over time can cause quantifiable modifications in the oxidation process of metals that could be harnessed for their use as a novel forensic tool.

36 MATERIALS SCIENCE↗
DOE OSTI2021

w19_OMEC Scientific Highlight [Slides]

Used DFT to identify likely experimental limitations for realizing high HER activity for DPA-based organic molecular electrocatalysts (OMEC) systems, directly guiding future experiments

08 HYDROGEN↗
DOE OSTI2021

Comparison of mechanical properties measured on multiple scales

FeCrAl alloys are being developed for accident tolerant fuels for current light water reactors. Specifically as fuel claddings with enhanced safety. These alloys are being developed to have excellent corrosion resistance in light water reactor coolant environments and structural integrity for longer durations in the case of a loss of coolant scenario where time is critical for safety. Tensile, shear punch, and Vickers hardness testing was performed on neutron irradiated FeCrAl alloys with three variations in composition. Testing was done at room temperature for comparison of the three techniques. FeCrAl alloys in this report were irradiated at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) to 7 dpa and at 320 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

BSU-8242 As-Run Thermal Analysis

This report contains the Engineering Calculations and Analysis Report (ECAR)-5510 documenting the thermal as-run analysis of the 1 and 3 DPA capsules that comprised the BSU-8242 Nuclear Science User Facilities Experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Mechanical and Thermophysical Properties of 3D-Printed SiC before and after Neutron Irradiation – FY21

This report presents the mechanical and thermophysical properties of 3D-printed SiC before and after neutron irradiation that have been evaluated to assess the fuel matrix material for the Transformational Challenge Reactor (TCR). The TCR fuel form consists of an additively-manufactured silicon carbide (SiC) matrix and uranium nitride tristructural isotropic (UN TRISO) fuel particles, which is manufactured through a newly developed processing route combining binderjet 3D printing, TRISO fuel particle loading, and chemical vapor infiltration (CVI). Because the fuel matrix is a primary component of the TCR core and its response to mechanical and thermal loads during operation is one of the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Baseline mechanical and thermophysical properties were measured from the disk specimens printed for different and sizes orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 280–310 MPa, which are slightly lower than that of the chemical vapor deposition (CVD) SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build direction of SiC samples, while thermal conductivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. Neutron irradiation tests were carried out on the 3D-printed 6-mm diameter SiC disk specimens. Irradiation was carried to 2.3 dpa over a temperature range of 360–880°C. No significant degradation in strength was observed in SiC after irradiations in various conditions and with different orientations. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

As-Run Physics Analysis for the EPRI Zirconium Growth C Capsule

This engineering calculation and analysis report documents the as-run nuclear-heating term, source-term, and radiation damage (measured in displacements per atom [dpa]) accumulation for the Electric Power Research Institute zirconium growth experiment's C capsule (EPRI-ZG-C). These data have been calculated using MCNP 6.1 and SCALE 6.2.3. These data can be used for: the shipping of the EPRI-ZG-C capsule, the acceptance of the capsule at the Materials and Fuels Complex, and programmatic scientific needs.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Cracking Behavior of Irradiated Ex-Plant Materials

The cracking behavior of reactor core internal materials is a key factor for the long-term safety and availability of light water reactors. Subject to high-temperature coolant, thermal and mechanical loading, and neutron irradiation, reactor core internal materials are susceptible to several degradation mechanisms during power operation. To understand the long-term effects of neutron irradiation, irradiated materials harvested from a decommissioned reactor were studied for their cracking susceptibility and fracture resistance. Small compact-tension specimens were machined from different locations on the baffle plates with irradiation doses ranging from 0.06 to 48.5 dpa. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed on these specimens in a simulated light water reactor environment, and the fracture morphologies of the tested samples were examined with a scanning electron microscope. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant load, no elevated cracking susceptibility was observed either in these samples at low stress intensity factors. However, a rapid cracking behavior featuring very high crack growth rates was observed in several high-dose samples. This rapid crack growth behavior was activated when the applied stress intensity factor, K, was sufficiently high. Once activated, the rapid crack growth remained steady even when the K was lowered to the pre-activation level and stopped only when the K was reduced drastically. Yet, the rapid crack growth could be re-activated repeatedly in these samples by raising the K above the initial activation level. This rapid cracking response may be related to the severe irradiation embrittlement experienced by these samples. The J-R curve tests performed with these samples showed a significant decline in fracture toughness with increasing neutron irradiation. Local cracking at the crack tip facilitated by the severe irradiation embrittlement of this material may be responsible for the rapid crack growth behavior demonstrated by this decommissioned material.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Experimental Evaluation of Deformation and Fracture Mechanisms in Highly Irradiated Austenitic Steels

The present report documents recent experimental results of analysis using scanning electron microscopy/electron backscatter diffraction (SEM-EBSD) of plastic deformation mechanisms and strain localization phenomena in austenitic steels irradiated by neutrons. Experiments were performed with specimens irradiated to 125 dpa and, additionally, with specimens that experienced radiation-induced swelling up to 3%. Section 1 briefly analyzes the deformation localization in irradiated steels and its consequences on the material performance. The section describes the advantages and importance of the SEM-EBSD approach combined with in situ mechanical testing capability. Section 2 briefly introduces the experimental tools and methods (i.e., SEM/EBSD in situ tensile frame, electric discharge machine to manufacture irradiated specimens) and describes the investigated materials (i.e., element composition, irradiation conditions, and general microstructure). Section 3 describes the key experimental results and provides a brief analysis and comparison with the datasets obtained earlier within the same task (i.e., low-dose specimens). The discussion focuses on EBSD microstructure maps with strain localization features, misorientation evolution as a function of strain, and observed deformation mechanisms. Section 4 evaluates data collected in recent years on highly irradiated steel and estimates the possible misorientation evolution under irradiation. The section introduces and discusses the concept of in-service-induced damage as an irradiation-assisted stress-corrosion cracking precursor. Section 5 summarizes the work performed. As expected, the present work results are beneficial for exploring and understanding degradation mechanisms in highly irradiated in-core materials found in light water reactors after long-term in-service life.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Status of the Optical Dilatometer Method of Evaluating the Peak Irradiation Temperatures of SiC Passive Monitors

The main objective of this project was to conduct a comparative assessment between the optical dilatometer method and resistivity method, using all 10 SiC temperature monitors provided by two Nuclear Science User Facility experiments: BSU-8242 and General Electric Hitachi. Unfortunately, due to multiple delays in acquiring, shipping, and cleaning the SiC temperature monitors, the project was only able to process one (1) SiC temperature monitor during this period. SiC temperature monitor KGT-3357 was evaluated via the optical dilatometer method to determine its peak irradiation temperature. The KGT-3357 sample was from the BSU-8242 experiment and designed for a temperature of 300°C and an exposure of 1 dpa. The optical dilatometer measurements indicated that the KGT-3357 SiC temperature monitor’s peak irradiation temperature range was 240–267°C, with sensitivity of approximately ±20°C. Additionally, this temperature range falls within the evaluated melt wire temperature range of 238.6–271.5 °C. The remaining six (6) SiC temperature monitors from the BSU-8242 experiment and three (3) from the General Electric Hitachi experiment will be used in the future work to further validate the optical dilatometer method for measuring SiC peak irradiation temperatures.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Report on Evolution of Inconel 718 Following HFIR Irradiation

The report presents the microstructure and mechanical properties of 3D printed Inconel 718 after irradiation in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) to assess its potential use as a structural material. The structural components near the outlets of several proposed reactor cores will experience significant neutron fluxes and outlet coolant temperatures ranging from the hot standby temperature of 300 °C to nearly 550 °C at the center of the part. These components must support the core in appropriate loading conditions and require structural analysis at relevant temperatures. In FY21, three heat treatments were designed and conducted to simplify the microstructure and to determine how each precipitating phase contributed to the overall strength. In FY21, baseline mechanical properties were measured from uniaxial tensile tests on subsize SS-J2 specimens at room temperature and at elevated temperatures of 300, 450, and 600 °C to serve a comparison to the irradiated properties. Four capsules containing 3D printed Inconel 718 were inserted into HFIR in FY21 for a matrix of two temperatures and two doses. The lower of the two doses was available for characterization in FY22. Multiple heat treatments of Inconel 718 irradiated to nominal conditions of 2 displacements per atom (dpa) at either 300 or 600 °C were strained with uniaxial tensile tests at the Irradiated Material Examination and Testing (IMET) Facility to discern the mechanical properties. Transmission electron microscopy was performed to correlate the observed mechanical properties with nanoscale features. The initially homogenous AM718-HM increased in strength at both irradiation temperatures based on a high density of nanometer-scale radiation-induced cavities at lower temperature and nucleation and growth of γ" precipitates at higher temperatures. The precipitate-hardened AM718-HT2 showed very small differences in strength before and after irradiation: the contribution to strength from γ" precipitates was replaced with dislocation loops. Because the radioactivity of the nickel superalloys from neutron activation limited the scope of the analysis, a feasibility study examined the possibility of an ultra-miniature specimen geometry, colloquially SS-Tiny (SS-T), for mechanical property determination using the nonirradiated Inconel 718. This study found an overestimation of ductility from the SS-T geometry with yield strength and ultimate tensile strength slightly above the SS-J2 geometry: this could be contributed to a reduction in specimen thickness.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ and ex-situ characterization of ion-irradiated AM materials

Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗