Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “IRRADIATION”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Second group of irradiation capsules: property data of irradiated welded stainless steel 347 and irradiated, welded, and hydrogen charged Zircaloy-4 for SHINE

Zircaloy-4 and AISI 347 stainless steel were irradiated to low doses (10 20 and 10 21 n/cm 2 ) at low temperature (~60°C) in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). These materials were under consideration for some of the structural components of the medical isotope production facility developed by SHINE Technologies. Zircaloy-4 base metal, as welded, welded and post-weld heat treated (PWHT), and hydrogen charged samples were irradiated. AISI 347 base metal from Sandmeyer Steel, Penn Stainless, and Rolled Alloys suppliers were irradiated. AISI 347 gas tungsten arc welding (GTAW) and flux-cored arc welding (FCAW) samples of Penn Stainless and Rolled Alloys were also irradiated. The irradiated materials were tested with microhardness and tensile tests at room temperature in air. The AISI 347 materials had more elongation after irradiation than the Zircaloy-4 materials. In fact, the Zircaloy-4 welded without PWHT and with H charging had the lowest elongation of any of the materials. This report is the final in a series of reports on this project.

36 MATERIALS SCIENCE↗
DOE OSTI2016

High-energy synchrotron study of in-pile-irradiated U–Mo fuels

Here synchrotron scattering analysis results on U–7wt.%Mo fuel specimens irradiated in the Advanced Test Reactor to three burnup levels (3.0, 5.2, and 6.3 × 1021 fission/cm3) are reported. Mature fission gas bubble superlattice was observed to form at intermediate burnup. The superlattice constant was determined to be 11.7 and 12.0 nm by wide-angle and small-angle scattering respectively. Grain sub-division takes place throughout the irradiation and causes the collapse of the superlattice at high burnup. The bubble superlattice expands the U–Mo lattice and acts as strong sink for radiation-induced defects. The evolution of dislocation loops was, therefore, suppressed until the bubble superlattice collapsed.

metallic nuclear fuel↗
DOE OSTI2021

Post-irradiation Examination Plan for the ORNL and University of California Santa Barbara Assessment of the UCSB ATR-2 Irradiation Experiment and a Reference Document for the Irradiated Archival RPV Materials Stored in the NSUF Nuclear Fuels and Materials Library

The Reactor Pressure Vessel task of the LWRS Program works with various organizations to obtain archival surveillance materials from commercial nuclear power plants to allow for comparisons of the irradiation-induced microstructural features from reactor surveillance materials with those from similar materials irradiated under high flux conditions in test reactors, such as the UCSB ATR-2 experiment. This report, originally submitted as a Level 3 Milestone M3LW-14OR0402012 – “Complete report on post-irradiation examination plan for ORNL and University of California Santa Barbara assessment of ATR-2 capsules,” has been expanded to providing a reference to access and / or perform characterization of irradiated archival RPV materials that were transferred to the Nuclear Science User Facility (NSUF), Nuclear Fuels and Materials Library (NFML).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiation-induced swelling and radiation-induced segregation & precipitation in dual beam irradiated Ferritic/Martensitic HT9 steel

Highlights: • The depth-dependence of radiation-induced cavities was characterized in self-ion irradiated HT9 co-implanted with He. • The injected interstitial effect was revealed to suppress cavity swelling in the vicinity of the damage peak. • ChemiSTEM elemental maps revealed the segregation of Ni at cavity surfaces and its possible synergistic influence on swelling was discussed. - Abstract: Ferritic/Martensitic HT9 steel was irradiated at 432 °C to 16.6 displacements per atom (dpa) (at the depth of 600 nm) using a defocused beam of 5 MeV Fe{sup ++} ions, while co-implanted with 3.22 appm He at the same depth. The helium concentration profile was designed so to follow the damage curve with a 0.22 appm He/dpa ratio at the depth from 300 to 1000 nm in the material. The depth-dependence of the cavity size and number density were characterized by both transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) imaging methods. A comparison between the two techniques was done showing good agreement. Cavity number density and the resulting swelling were found to be suppressed by the injected interstitial effect in the vicinity of the ion induced damage peak. The region between 300 and 750 nm depth which excludes the injected interstitial effect was thus proposed for improved cavity swelling analysis. The swelling ratio in this region was found to be ~(0.86–1.02) × 10{sup − 2}%/dpa. In addition, ChemiSTEM characterization revealed radiation-induced segregation occurring throughout the irradiated region and precipitation of G-phase particles. Segregation of Ni to cavity surfaces was also observed and its possible synergistic influence on swelling was discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Microstructure Evolution in Ion-Irradiated Oxidized Zircaloy-4 Studied with Synchrotron Radiation Microdiffraction and Transmission Electron Microscopy

The corrosion process (oxidation and hydriding) of zirconium alloy fuel cladding is one of the limiting factors on fuel rod lifetime, particularly for Zircaloy-4. The corrosion rate of this alloy shows indeed a great acceleration at high burnup in light water reactors (LWRs). Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. In this study, Zircaloy-4 samples underwent helium and proton ion irradiation up to 0.3 dpa, forming a uniform defect distribution up to 1 pm deep. Both as-received and precorroded samples were irradiated to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples were corroded to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray microdiffraction and microfluorescence were used to follow the evolution of oxide crystallographic phases, texture, and stoichiometry both in the metal and in the oxide. In particular, the tetragonal oxide phase fraction, which has been known to play an important role in corrosion behavior, was mapped in both unirradiated and irradiated metals at the submicron scale and appeared to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest, were combined to fully characterize changes caused by irradiation in metal and oxide phases of both alloys.

ion-irradiation↗
DOE OSTI2017

Modeling thermal and irradiation-induced swelling effects on the integrity of Ti 3 SiC 2 /SiC joints

Previously, results for CVD-SiC joints created using solid state displacement reactions to form a dual-phase SiC/MAX phase irradiated at 800°C and 5 dpa indicated some extent of cracking in the joint and along the CVD-SiC/joint interface. This paper elucidates the origin of cracking by thermomechanical modeling combined with irradiation-induced swelling effects using a continuum damage approach with support of micromechanical modeling. Three irradiation temperatures (400°C, 500°C and 800°C) are considered assuming experimental irradiation doses in a range leading to saturation swelling in SiC. The analyses indicate that a SiC/MAX joint heated to 400°C fails during irradiation-induced swelling at this temperature while it experiences some damage after being heated to 500°C and irradiated at the same temperature. However, it fails during cooling from 500°C to room temperature. The joint experiences minor damage when heated to and irradiated at 800°C but does not fail after cooling. The prediction agrees with the experimental findings available for this case.

damage modeling↗
DOE OSTI2022

Emulation of neutron-irradiated microstructure of austenitic 21Cr32Ni model alloy using dual-ion irradiation

Here, in this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors. The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at ~380°C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430°C-500°C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460°C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at ~380°C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460°C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation.

21Cr32Ni model alloy↗
DOE OSTI2016

Thermophysical property and pore structure evolution in stressed and non-stressed neutron irradiated IG-110 nuclear graphite

The nuclear graphite, IG-110, was irradiated with and without a compressive load of 5 MPa at ~400 *C up to 9.3E25 n/m2 (E > 0.1 MeV). Following irradiation physical properties were studied to compare the effect of graphite irradiation on microstructure developed under compression and in stress-free conditions. Properties included: dimensional change, thermal conductivity, dynamic modulus, and CTE. The effect of stress on open internal porosity was determined through nitrogen adsorption. The IG-110 graphite experienced irradiation-induced creep that is differentiated from irradiation-induced swelling. Irradiation under stress resulted in somewhat greater thermal conductivity and coefficient of thermal expansion. While a significant increase in dynamic modulus occurs, no differentiation between materials irradiated with and without compressive stress was observed. Nitrogen adsorption analysis suggests a difference in pore evolution in the 0.3e40 nm range for graphite irradiated with and without stress, but this evolution is seen to be a small contributor to the overall dimensional change.

Snead, Lance↗
DOE OSTI2017

In Situ TEM Multi-Beam Ion Irradiation as a Technique for Elucidating Synergistic Radiation Effects

Materials designed for nuclear reactors undergo microstructural changes resulting from a combination of several environmental factors, including neutron irradiation damage, gas accumulation and elevated temperatures. Typical ion beam irradiation experiments designed for simulating a neutron irradiation environment involve irradiating the sample with a single ion beam and subsequent characterization of the resulting microstructure, often by transmission electron microscopy (TEM). This method does not allow for examination of microstructural effects due to simultaneous gas accumulation and displacement cascade damage, which occurs in a reactor. Sandia’s in situ ion irradiation TEM (I3TEM) offers the unique ability to observe microstructural changes due to irradiation damage caused by concurrent multi-beam ion irradiation in real time. This allows for time-dependent microstructure analysis. A plethora of additional in situ stages can be coupled with these experiments, e.g., for more accurately simulating defect kinetics at elevated reactor temperatures. This work outlines experiments showing synergistic effects in Au using in situ ion irradiation with various combinations of helium, deuterium and Au ions, as well as some initial work on materials utilized in tritium-producing burnable absorber rods (TPBARs): zirconium alloys and LiAlO2.

Taylor, Caitlin↗
DOE OSTI2019

Local structure of Ni 80 X 20 (X: Cr, Mn, Pd) solid-solution alloys and its response to ion irradiation

In this work, the local structure of Ni 80 X 20 (X: Cr, Mn, Pd) solid-solution alloys was investigated with x-ray absorption and total scattering x-ray diffraction methods. Atomic pair distribution function (PDF) analysis indicated that the local lattice distortion is strongly relevant to the atomic size mismatch, and the local lattice distortion in Ni80Pd20 alloy is obviously larger than that in other solid-solution alloys. The bond length of different atomic pairs was derived from the fitting of extended x-ray absorption fine structure spectra. Quantitative analysis of the local bonding environment in Ni 80 Cr 20 during Ni ion irradiation suggested that Cr atoms tend to form clusters in Ni 80 Cr 20 with the increase of ion dose.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, T o , in alloys irradiated to 7 dpa and higher.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Irradiation performance of AGR-1 high temperature reactor fuel

The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110m Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10 –4 to 5 × 10 –4 for 154 Eu and 8 × 10 –7 to 3 × 10 –5 for 90 Sr. The average 134 Cs fractional release from compacts was <3 × 10 –6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10 5 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134 Cs fractional release in two capsules to approximately 10 –5 . Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization of these elements within the SiC microstructure is the subject of ongoing focused study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Key results from irradiation and post-irradiation examination of AGR-1 UCO Triso fuel - 18576

The AGR-1 irradiation experiment was performed as the first test of tri-structural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 x 10{sup 5} coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136 deg. C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 deg. C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Deformation mechanism study of a hot rolled Zr-2.5Nb alloy by transmission electron microscopy. II. In situ transmission electron microscopy study of deformation mechanism change of a Zr-2.5Nb alloy upon heavy ion irradiation

The effect of heavy-ion irradiation on deformation mechanisms of a Zr-2.5Nb alloy was investigated by using the in situ transmission electron microscopy deformation technique. The gliding behavior of prismatic 〈a〉 dislocations has been dynamically observed before and after irradiation at room temperature and 300 °C. Irradiation induced loops were shown to strongly pin the gliding dislocations. Unpinning occurred while loops were incorporated into or eliminated by 〈a〉 dislocations. In the irradiated sample, loop depleted areas with a boundary parallel to the basal plane trace were found by post-mortem observation after room temperature deformation, supporting the possibility of basal channel formation in bulk neutron irradiated samples. Strong activity of pyramidal slip was also observed at both temperatures, which might be another important mechanism to induce plastic instability in irradiated zirconium alloys. Finally, (011{sup ¯}1)〈01{sup ¯}12〉 twinning was identified in the irradiated sample deformed at 300 °C.

36 MATERIALS SCIENCE↗