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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

STEM imaging of irradiation induced defects in Eurofer97 steel variants irradiated in the EUROfusion collaboration

Ten exploratory variants of Eurofer97 reduced activation ferritic martensitic (RAFM) steel were irradiated in HFIR to ~2.94 – 32.4 dpa, 300±30 °C as a part of the EUROfusion collaboration. The irradiations were performed in rabbit capsules. ES21-22 capsule were used for irradiating SS-J3 tensile samples while M4CVN bend bar samples were irradiated in ES31-35 capsules. The mechanical properties and multi-length scale characterization of the microstructures of these ten steels, designated as H, I, P, J, K, L, M, N, O, P and reference E, in the nonirradiated form was reported in "Mechanical properties and microstructure characterization of unirradiated Eurofer-97 steel variants for the EUROfusion project, ORNL/SPR-2018/882," while their irradiated properties and some microstructure analysis was reported in "Post-irradiation examination of Eurofer97 steel variants irradiated to 2.5 dpa, ~300 °C in HFIR for the EUROfusion program, ORNL/SPR-2020/1440." Here, additional microstructures using TEM/STEM and STEM-EDX of the irradiated steels is presented. Characterization was performed at the Low Activation Materials Development and Analysis (LAMDA) laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Irradiation And Post-Irradiation Examination Plan For A709

The Advanced Materials and Manufacturing Technologies program is proposing a neutron irradiation campaign for alloy 709 (A709) plate product developed by the Advanced Reactor Technologies program. This irradiation campaign will support the deployment of A709 (included welded components) by providing data of engineering importance, such as data that can be used for establishing end-of-life design parameters and knockdown factors for irradiated materials properties versus properties for non-irradiated material. It will also advance the scientific understanding of A709 irradiation behaviors and material property evolution, which is vital for deploying A709 components into reactors with different neutron spectra, neutron fluxes, and irradiation temperatures as well as accelerating the deployment of new product forms and weld procedures. This document presents the plan for the neutron irradiation and post-irradiation examination for A709 and the scientific and technical basis for the plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1990

An Estimate of Changes in the Sun's Total Irradiance Caused by UV Irradiance Variations from 1874 to 1988

Enhanced emission from bright solar faculae is a source of significant variation in the sun's total irradiance. Relative to the emission from the quiet sun, facular emission is known to be considerably greater at UV wavelengths than at visible wavelengths. Determining the spectral dependence of facular emission is of interest for the physical insight this may provide to the origin of the sun's irradiance variations. It is also of interest because solar radiation at lambda less than 300 nm is almost totally absorbed in the Earth's atmosphere. Depending on the magnitude of the UV irradiance variations, changes in the sun's irradiance that penetrates to the Earth's surface may not be equivalent to total irradiance variations measured above the Earth's atmosphere. Using an empirical model of total irradiance variations which accounts separately for changes caused by bright faculae from those associated with dark sunspots, the contribution of UV irradiance variations to changes in the sun's total irradiance is estimated during solar cycles 12 to 21.

Lean, J.↗
DOE OSTI2020

A quantitative study of retention and release of deuterium and tritium during irradiation of y-LiAlO2 pellets

This study reports on the quantification of deuterium in ion-irradiated gamma-LiAlO2 pellets as a function of dose and temperature. The gamma-LiAlO2 pellets were sequentially irradiated with He+ and D2+ ions to the same fluences of 5E16, 1E17 and 2E17 He++D+/cm2 at 188 K. Additional irradiation was performed to 1E17, 2E17 and 4E17 He++D+/cm2 at 573 K. A set of the pellets irradiated at 188 K was shipped and stored at low temperatures from 80 to 132 K and characterized using time-of-flight secondary ion mass spectrometry at ~173 K. The deuterium depth profiles show a Gaussian-like distribution in the low-temperature pellets. The total deuterium retention is found to be directly proportional to the ion fluence. About 27 at.% of the implanted deuterium atoms were released from the pellet irradiated to 2E17 He++D+/cm2 at 188 K during storage at room temperature for ~1 month. Retention of the trapped or bound deuterium during ion irradiation at 573 K increases initially with ion fluence and tends to saturate at a high fluence. The amount of the released deuterium is observed to be quantitatively consistent with that of the released tritium from similar standard pellets during neutron irradiation at 573 K.

Deuterium retention and release, ion irradiation, ↗
DOE OSTI2021

Fission product distribution in irradiated safety-tested and as-irradiated AGR-2 TRISO particles

In this work, two tristructural isotropic (TRISO)-coated nuclear fuel particles were examined by electron probe microanalysis (EPMA) as part of the Advanced Gas Reactor program. The compacts’ average irradiation temperatures ranged from approximately 1260 to 1290 °C. One particle was examined in the as-irradiated condition, while the other was subject to 1600 °C post-irradiation safety testing. This study was undertaken to test a newly-developed EPMA technique to determine fission product masses in TRISO particles on a layer-by-layer basis, and to compare fission product distributions between an as-irradiated and safety-tested particle. Fission product concentration profiles were collected along two radii in each particle, with measured concentrations used to compute the fission product mass in each TRISO particle layer. These measured masses were then compared to those predicted from ORIGEN modeling calculations. Data collected from these measurements show that for these two particles, masses determined via EPMA were within ± 20% of the calculated masses for the rare-earth elements, Mo, Zr, Cs, I, and Pd. Elements that tend to be less homogeneously distributed include Sr, Te, Eu, Ag, and possibly Ba. Measured Ag masses differed by more than 40% from the calculated mass. Lanthanides other than Eu remain primarily within the fuel kernel in the as-irradiated particle but in the safety-tested particle these element masses were divided approximately equally between the kernel and kernel periphery. In both particles, the majority of Sr and Eu accumulated in the carbon-rich kernel periphery, although in the safety-tested particle, Sr and Eu accumulated farther from the fuel kernel than occurred with irradiation alone. A greater mass fraction of mobile elements, such as Cs and I accumulated in the buffer and IPyC in the safety-tested particle as compared to the as-irradiated particle. When fully developed and tested, this mass balance approach to TRISO particle analysis has the potential to provide insight into fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of a MOOSE-based crystal plasticity model with irradiation defect evolution for irradiation creep in 316

Irradiation creep and irradiation swelling are two of the lifetime limiting factors for structural materials in nuclear reactors. These mechanical effects are driven by irradiation defect evolution and the interaction of those defects with dislocations in the microstructure. We present here a coupled cluster dynamics – crystal plasticity approach to model irradiation swelling and creep behavior in 316 SS. The time-dependent evolution of irradiation defects is calculated with a cluster dynamics approach and passed to the crystal plasticity model to compute the dislocation evolution. We show the impact of the irradiation defect evolution on the stress state in the material, which drives inelastic deformation through dislocation motion. The inelastic deformation in the 316 SS is dependent on the dose rate, where the inelastic deformation driven by the early-stage irradiation defect evolution determines the mechanical behavior of the 316 SS.

316 Stainless Steels↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Comparison of the Changes in the Visible and Infrared Irradiance Observed by the SunPhotometers on EURECA to the UARS Total Solar and UV Irradiances

Solar irradiance in the near-UV (335 nm), visible (500 nm) and infrared (778 nm) spectral bands has been measured by the SunPhotometers developed at the World Radiation Center, Davos, Switzerland on board the European Retrievable Carrier between August 1992 and May 1993. Study of the variations in the visible and infrared irradiance is important for both solar and atmospheric physics. The purpose of this paper is to examine the temporal variations observed in the visible and infrared spectral bands after eliminating the trend in the data mainly related to instrument degradation. The effect of active regions in these spectral irradiances is clearly resolved. Variations in the visible and infrared irradiances are compared to total solar irradiance observed by the SOVA2 radiometer on the EURECA platform and by the ACRIMII radiometer on UARS as well as to UV observations of the UARS and NOAA9 satellites. The space-borne spectral irradiance observations are compared to the photometric sunspot deficit and CaII K irradiance measured at the San Fernando Observatory, California State University at Northridge in order to study the effect of active regions in detail.

EURECA UARS↗
DOE OSTI2022

Cryogenic tracer irradiation facility at the university of Texas at Austin

We report on a cryogenic gas irradiation facility in the 1.1 MW TRIGA reactor at The University of Texas at Austin. The system was designed to produce radioactive xenon and argon for environmental studies, and it can be applied to produce other gaseous radiotracers. The system design includes modeling accident scenarios to ensure operation does not risk damage to the reactor and minimizes risk of release of radioactive material. In conclusion, the SCALE code was used to model a 30-day irradiation of one liter of Xe-126 and the predicted activity is 0.9 Ci of Xe-127 at irradiation end.

Ar-37↗
DOE OSTI2025

Irradiation-induced formation of G-phase precipitates and M 2 X carbides in self-ion irradiated HT-9

Ferritic-martensitic steels with high chromium content are a promising material group for advanced nuclear systems due to their high temperature strength and good irradiation tolerance. HT-9 is an optimized and often-studied alloy in this group, but additional studies are required on its radiation response under extreme conditions to be experienced in various types of nuclear reactors, especially with respect to phase stability under irradiation. Self-ion irradiation of HT-9 by 5 MeV Fe ions was used to simulate neutron-induced behavior reaching peak doses of 100 and 300 dpa at temperatures ranging from 450 to 550 °C. M 23 C 6 carbides that existed prior to irradiation were found to remain stable under all examined irradiation conditions. As irradiation progressed at 450 and 500 °C, however, formation of spherical-like G-phase precipitates and needle-like M 2 X carbides was observed. G-phase precipitates were found to be enriched in Ni, Si, and Mn, and show no interface segregation, whereas needle-like M 2 X carbides were rich in Cr and Mo and clearly displayed interface segregation of Ni and Si. M 2 X carbide formation is believed to be assisted by vacancies, while G-phase precipitation is thought to be assisted by interstitials. Finally, this difference in defect-mediated formation leads to a difference in distribution with depth. M 2 X carbides are distributed over shallower depths than that of G-phase precipitates, consistent with defect imbalance predictions that consider the influence of the injected interstitial effect.

atom probe tomography↗
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↗
DOE OSTI2024

Post-Irradiation Examination to Quantify Irradiation-Induced Bowing of SiGA® Silicon Carbide Composite Structures (Final CRADA Report – NFE-23-09937)

As part of the SiC-based material development at General Atomics Electromagnetic Systems (GA-EMS), this project involved the first-of-a-kind experimental post-irradiation examination of the irradiation-induced bowing response in SiC composite structures under a neutron flux gradient. The SiC composite miniature channel specimen was provided by GA-EMS for the irradiation experiment. To quantify the irradiation-induced bowing of the channel specimen, a series of visual and dimensional inspections were conducted using the unique capabilities at Oak Ridge National Laboratory (ORNL), including a custom profilometry rig. The post-irradiation examination results of this project will assist GA-EMS in validating their fuel performance model for SiC-based core structures in neutron irradiation environments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

FAST irradiations and initial post irradiation examinations – Part I

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. Tests included the use of fuel additives (Sn, Sb, and Pd) for reducing the gettering of lanthanide fission products, zirconium liners to mitigate fuel-cladding chemical interaction, and annular fuel geometry for sodium-free fuel designs. The first cycle of FAST testing was completed and a series of rodlets were removed for the initial post irradiation examination (PIE) of FAST rodlets. The rodlet irradiation conditions was evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % heavy metal [%FIMA]) control rodlets and a helium bonded annular rodlet (4.7 %FIMA). Initial non-destructive PIE has been completed and includes, visual inspection, neutron radiography and gamma scanning of the FAST rodlets. Radiography confirmed the integrity of the experiments as well as showing the complete filling of the annulus in the annular fuel at a modest burnup (4.7 %FIMA) and potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior with the exception of the ceasium in the He-bonded annular fuel. Future destructive PIE will be necessary in order to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Irradiation of an enriched uranium (NaCl-UCl 3 ) fuel salt capsule, summary of nondestructive post-irradiation examinations, and solidification modeling

Molten salt reactors (MSRs) are gaining attention due to their potential for safe, carbon-free nuclear energy with reduced waste. However, licensing these reactors is hindered by limited experimental data on fueled salts, both pre- and post-irradiation. Here, the novel Molten-salt Research Temperature-controlled Irradiation (MRTI) vehicle was designed to address knowledge gaps in irradiating enriched‑uranium-bearing salts. The MRTI experiment irradiated 13 cm 3 of UCl₃-NaCl (93 % U-235) salt in the Neutron Radiography (NRAD) Reactor, achieving a burnup of 0.196 GWd/MTU over 390 h. Despite a heater failure, the thermocouple data suggested fission heat kept the salt molten. The MRTI assembly was remotely disassembled for nondestructive post-irradiation examination (PIE), which included precision gamma-ray scanning (PGS) and neutron radiography. Radiograph images showed the location of the salt and the solidification pattern. PGS results provided an early indication that activated materials of construction did not increase in relative intensity in the region of the capsule where the salt was in contact with the material of construction. Additionally, PGS data showed the presence of several gamma emitting fission products, such as Nb-95, Zr-95, Ru-103, Ce-141, and La-140, where Ru-103 had the highest counts at the bottom of the capsule. Computational fluid dynamics modeling supported observations of salt solidification patterns and proved to be a valuable tool to inform PIE activities. The MRTI experiment has thus far provided critical data and lessons learned for fuel salt PIE activities, essential for advancing the technical readiness of MSRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comparative Mechanical Properties Analysis of Triple Ion-Beam Irradiated and Neutron Irradiated Potential Plasma Facing Components

Abstract Several classes of materials are being proposed for use in fusion reactors including oxide dispersion strengthened (ODS) and reduced activation ferritic-martensitic (RAF/M) steels to withstand the severe and harsh conditions. In this work, the mechanical properties of a Fe-16Cr-4Al-2W-0.3Ti-0.3Y 2 O 3 (K3) (ODS) ferritic steel and a Fe-8.9Cr-1.1W-0.47Mn-0.2V-0.14Ta-0.11C (Eurofer 97) (RAF/M) steel) after triple ion beam irradiation were locally evaluated utilizing in-situ micro-pillar compression tests, and continuous stiffness/quasi-static nanoindentation. No change in mechanical properties was observed in the K3 ODS steel. However, the Eurofer 97 RAF/M steel exhibited radiation-induced effects via increases in yield strength. Micro-pillar techniques were expanded to neutron-irradiated materials via an in-situ testing technique employing lift-out methods on Fe-14Cr-0.9Ti-0.3Mo-0.25Y 2 O 3 (MA957) ODS ferritic steel. Both the non-irradiated and irradiated compressive yield stresses of the MA 957 micro-pillars were in good agreement with bulk yield stress values reported in the literature, suggesting that the lift-out micro-pillar compression testing technique is a promising method. The demonstration of these techniques on ion beam and neutron irradiated ODS steels and ion beam RAF/M steels gives information to inform models of the material degradation during use in a fusion device.

alloys↗
DOE OSTI2020

Investigation of the irradiation effects in additively manufactured 316L steel resulting in decreased irradiation assisted stress corrosion cracking susceptibility

While additive manufacturing (AM) is an area of increasing interest to the nuclear industry, there are significant research and development needs prior to successful implementation of additively manufactured components into reactors. AM steel behaves differently than traditional wrought material, as has been observed in numerous mechanical testing studies. The nuclear industry must also understand the material response to irradiation. This work examines the irradiation damage and the irradiation assisted stress corrosion cracking behavior (IASCC) of both proton irradiated AM and wrought 316L steel. For IASCC testing, specimens were strained to 4% plastic strain in a simulated boiling water reactor environment (288°C, 0.2 µS/cm). Transmission and scanning electron microscopy, confocal microscopy, and X-ray computed tomography were used to examine coupons of the material as well as tensile bars used in the IASCC testing. This work found that the wrought 316L was more susceptible to IASCC than the AM steel, however, some IASCC was observed in the AM steel when tested with the tensile axis parallel to the build direction. Electron microscopy has shown significantly fewer radiation induced voids formed in the AM steel. While the pre-existing pore network in the AM steel may act as a sink to point defects, this is not expected to be the major factor in the reduced swelling observed in the AM steel. The lower density of radiation induced voids was expected to contribute to the IASCC resistance by decreasing the localization of deformation in the irradiated specimens, however, severity of localized deformation in AM steel, measured in dislocation channel spacing and height, was similar or more severe to that found in wrought. In conclusion, the presence of severe cracking in the wrought specimens did not allow for a direct comparison of localized deformation severities between the AM and wrought steel.

Additive manufacturing↗