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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Addendum to Capability Needs for Irradiated and Radioactive Materials Research (Ad Hoc Committee Summary Report)

Nuclear materials and fuels studies are challenging research and engineering targets because of their inherent radioactivity as well as their heterogeneous microstructure. Advanced light sources have made significant underpinning scientific contributions to the understanding of structural and fuel cladding material microstructure and properties and the development of structure-property relationships. These facilities have the potential to help advance the Department of Energy (DOE) Office of Nuclear Energy (DOE-NE) mission priorities providing underpinning understanding of technically important challenges, including: irradiation induced embrittlement and swelling; stress corrosion cracking and corrosion in extreme environments; ageing of reactor pressure vessels steels; nuclear fuel characterization; and nuclear waste form optimization as well as playing a major role in the development of a materials performance matrix to aid in the design of new materials to meet the needs of advanced reactor concepts. However, the delivery of underpinning understanding is the mission of DOE Office of Science (DOE-SC). While filling such knowledge gaps is important, efforts in this direction should not be confused with the strategic technology focused mission goals of DOE-NE to enable the continued operation of existing U.S. nuclear reactors, to enable the deployment of advanced nuclear reactors, to develop advanced nuclear fuel cycles, and to maintain U.S. leadership in nuclear energy technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Accelerating Nuclear Fuels and Materials Qualification by Multi-Level Irradiation Experiment Campaign

The advanced reactor technologies feature fuels, coolants, and materials that promise safer operating conditions under normal and accident scenarios. However, the nuclear fuels and materials qualifications require several decades (for example, new reactor fuel qualification from conceptualization requires about 20 years). Therefore, accelerating nuclear fuels and materials qualification is essential, and it can be achieved by combining high through-put materials irradiation and testing, advanced post-irradiation examinations, and Multiphysics modeling. This paper addressed the associated challenges in accelerating nuclear fuels and material qualification for new and advanced reactor designs, which differ based on fuel, coolant, operating conditions, and structural materials. These challenges vary for radiation level, operating conditions (e.g., temperature and pressure), and coolant type (e.g., corrosion environment). In addition, the challenges and limitations in modeling tools, experimental facilities, and licensing guidelines are also discussed, and a general solution path forward is recommended.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Presentation: Accelerating Nuclear Fuels and Materials Qualification by Multi-Level Irradiation Experiment Campaign

The advanced reactor technologies feature fuels, coolants, and materials that promise safer operating conditions under normal and accident scenarios. However, the nuclear fuels and materials qualifications require several decades (for example, new reactor fuel qualification from conceptualization requires about 20 years). Therefore, accelerating nuclear fuels and materials qualification is essential, and it can be achieved by combining high through-put materials irradiation and testing, advanced post-irradiation examinations, and Multiphysics modeling. This paper addressed the associated challenges in accelerating nuclear fuels and material qualification for new and advanced reactor designs, which differ based on fuel, coolant, operating conditions, and structural materials. These challenges vary for radiation level, operating conditions (e.g., temperature and pressure), and coolant type (e.g., corrosion environment). In addition, the challenges and limitations in modeling tools, experimental facilities, and licensing guidelines are also discussed, and a general solution path forward is recommended.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Crack Growth Rate and Fracture Toughness Tests on Irradiated Ex-Plant Materials

The performance of structural materials is critical for the safe and economic operation of light water reactors. Exposed to neutron irradiation during service, the reactor core internal materials can undergo significant microstructural and microchemical changes, leading to irradiation hardening and embrittlement. To ensure the structural integrity and functionality of nuclear reactor components during long-term operation, material degradation and damage mechanisms must be understood and managed adequately. In this work, irradiated materials harvested from the decommissioned Zorita reactor were studied for their cracking susceptibility and fracture resistance as a function of irradiation dose up to 47 displacement per atom (dpa). The material is a Type 304 stainless steel sectioned from the baffle plates of this pressurized water reactor with 38 years of service. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed in low-corrosion-potential environments at ~315°C. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant stress intensity factors, most samples did not show elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance in the test environments. However, an unstable cracking behavior was observed in a 47-dpa sample, resulting in significantly higher crack growth rates than expected at high stress intensity factors. Crack instability was also observed in a 0.06-dpa sample but did not lead to a sustained high crack growth rate. The impact of neutron irradiation was more evident in the fracture toughness J-R curve tests. As the dose increased, the J-R curve declined considerably and became very shallow at high doses. A fully intergranular fracture morphology was also observed among the high-dose samples ruptured in an air atmosphere at room temperature. This brittle fracture mode in the absence of high temperature water environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Femto-second Laser’s Enabling New Length Scale Fabrications for Rapid Post Irradiation Examination of Materials: Concluding LDRD Project Poster

Mechanical testing campaigns are required to qualify materials for advanced reactor conditions, yet economical and safety limitations restrict the number of standardized mechanical tests that can be performed. Reducing the size of the sample is one approach to addressing these challenges and to accelerating testing. Previous research has shown that smaller mechanical test samples produce higher yield and ultimate stress values compared to values measured from standard sample sizes: the “smaller is stronger” effect. Specimens used in accelerated material testing campaigns must reflect bulk material performance to enable engineering scale material property measurement. The objective of this research project was to determine if engineering scale mechanical behavior—the yield stress—could be measured with micro-tensile test samples smaller than traditional standard testing geometries. The relationship between yield stress and sample size was explored with two different nuclear-relevant structural materials: Zircaloy-4 and tungsten. Mechanical testing of both metals demonstrated decreasing yield stress values with increasing sample gauge size across three different sizes. Yield stress values from the largest gauge size, fabricated with a femto-second laser ablation system, approach bulk material yield stress values reported in published literature. Preliminary analysis of the tungsten samples indicates the yield stress value depends on the grain characteristics within the gauge section, in addition to the gauge size. Accompanying modeling efforts, including response surface generation and crystal plasticity approaches, further demonstrated that the size of the sample gauge section alone cannot explain the change in yield stress values.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Irradiation Testing of Additively Manufactured Materials for LWR Applications: Alloy 718 and 316L Stainless Steel

The objective of this NSUF Project is to assess the changes in irradiated additively manufactured (AM) material properties as compared to non-irradiated material. Type 316L stainless steel and Alloy 718 samples were produced using Direct Metal Laser Melting (DMLM) fabrication. Materials produced from this fabrication method have several potential applications within the nuclear industry as reactor internal repair parts, fuel debris resistant filters, or fuel spacers within existing light water reactors (LWRs). AM materials have been shown to achieve equivalent mechanical behavior in simulated reactor environments as compared to wrought materials, but have significantly more flexibility when it comes to unique design features. The increased component design flexibility makes these AM materials an attractive choice for both current LWR applications as well as for small modular reactor (SMR) designs. Prior to use of these materials in reactor fleet operation, the industry as a whole must evaluate the effects of irradiation on their material properties. Standard 0.4 inch thick Compact Tension specimens and SSJ3 type tensile bars were neutron irradiated at the Advanced Test Reactor to ~1 dpa for the purpose of performing a variety of mechanical tests in a range of simulated environments applicable to LWRs. For the ductile austenitic Type 316L stainless steel, the irradiated data will be used to confirm that the AM process produces materials with properties that are equivalent to wrought materials under testing conditions applicable to LWR operation. Transmission electron microscopy analysis was also performed in order to understand microstructural and microchemical changes induced in each material in response to neutron irradiation. If possible, data collected from these AM 316L samples will be used to remove fluence limits from specifications of ASME code cases for this alloy, which will give vendors much more flexibility in building future components.

36 - MATERIALS SCIENCE↗
DOE OSTI2023

Quick Look Report of Godiva U233 shot March 2022 Data Collection Activities

This report discussing the March 2022 irradiation of U233 on the Godiva critical assembly using burst mode and subsequent analysis of the fission products starting at 65 minutes post irradiation, and ending 12 days later. The Short-Lived Fission Product Yield (SLFPY) project is one of many efforts under the Nuclear Physics Multi-Lab and related Venture Project. The objective of the SLFPY project is to provide improved measurements of fission product yields for select nuclides. The SLFPY project consists of measurement and analysis campaigns. To produce fission data, samples of actinide material are irradiated in a high-flux neutron environment. These samples are quickly placed in front of a detection system where both gamma singles and coincidence spectra are collected as a function of time. Gamma spectroscopy is performed on the irradiated material on site near the irradiation facility, starting as soon as feasible after the irradiation and continuing for 7 days or longer. Following the data collection, time-dependent fission yield analyses are performed using gamma-ray spectroscopy. This document summarizes the measurement campaign performed March 21 - April 4, 2022 where sub-gram quantities of 233 U nitrate (UO 2 (NO 3 ) 2 ) were irradiated in the Godiva reactor. The Godiva burst #2076 conducted under experiment IER 504 had a temperature change of +156 degrees Celsius, lasted 13.32 microseconds, and had a reactivity of $\$$1.089. The targets consisted of uranium oxide solution deposited in a quartz tube and packed with quartz wool. The flame-sealed quartz ampoules were used to contain the uranium samples for the irradiation and count activities. On March 22, 2022, at 10:27:00 a.m. the three uranium ampoules, along with a witness foil pack, were irradiated in the Godiva critical assembly using burst mode operations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Overview/Recap of the 2020 Accelerated Irradiation of Reactor Structural Materials Workshop

The workshop provided a snapshot look at current international, in-reactor testing and irradiation capabilities. Gaps were identified where possible, and discussions focused on potential mitigation strategies and recommended paths forward. As new material innovations are being developed for in-reactor applications for reactor life extension long-term operation, and advanced reactor technologies, there is an increased need for materials qualification and assessment programs. In-reactor testing capabilities are vital to the on-going success of these DOE-NE programs and initiatives. In recent years, multiple international research reactors have been placed in a permanent shut-down state and have begun full decommissioning. These activities have created new risks in the delivery of on-going and/or expected research programs. Beyond performing simple irradiations in test reactors, few facilities exist internationally which can perform instrumented, in-situ irradiations on structural materials, and with the recent shut down of facilities hosted at Halden and NRU (example: instrumented fatigue loop, and in-situ creep), there are further gaps in the industry left un-filled. A new focus is being placed on the use of accelerator- based technologies to fill in some of these gaps, but these must be viewed as supplemental, and not surrogates to in-reactor capabilities

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe 3+ and 3.42 MeV energy degraded He 2+ ) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. Furthermore, the comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Micromechanical properties of spherical and facetted He bubble loaded copper

Exploring new irradiation resistant materials requires understanding their mechanical responses to irradiation. Resistance to helium bubble formation and understanding bubble effects on the mechanical response of candidate materials are crucial factors to qualify materials as irradiation resistant. Here, in this paper, we explore the effect of spherical and facetted helium bubbles on the mechanical response of copper via in-situ micromechanical tensile testing at room temperature. Bubble formation and shape effects on strength and ductility, and their behavior on grain boundaries are discussed and compared to literature. Loading Cu with helium bubbles is shown here to increase strength but decrease ductility.

42 ENGINEERING↗
DOE OSTI2024

Dynamic Scaling Analysis of Accelerated Irradiation Testing on Additive Manufacturing Materials by Positron Annihilation

The timely applications of Additive Manufacturing (AM) materials in nuclear environments require accelerated irradiation tests, mainly ion irradiation to enable rapid prototyping. Low dose ion irradiation would cause sub-nanostructure changes by generation of lattice defects, vacancies, vacancy clusters and voids and void swelling caused by cellular dislocations. Positron Annihilation Lifetime (PAL), a novel technology, sensitive towards sub-nanostructure morphology with high accuracy (about 10-7 vacancy per atom), supported by Transition Electron Microscope (TEM) would be applied to identify the type and total size of the defects. The subsequent PAL measurements and TEM surface studies would be followed by PAL analysis that includes sophisticated trapping model. The PAS results would become an input to dynamic scaling analysis (that predicts radiation effects from low dose studies for high dose effects), which incorporate mean-field theory model. The final effect is an in-depth understanding of the microstructure evolution of AM materials under ion irradiation which can be extrapolated to the studies of neutron irradiation, since ion-irradiation takes less time and do not cause the irradiation hazard. The working hypothesis is that PAL technology, that have excellent sensitivity to low-defect concentration would help to identify ion-induced material damage on the atomic and nano-scale level, which then could be extrapolated to understand the neutron damage better.

accelerated irradiation testing↗
DOE OSTI2025

Shielded magnetic small-angle neutron scattering for characterization of radioactive samples

The development of a Pb-shielded fixture for the execution of a small-angle neutron scattering (SANS)-based workflow for interrogation of highly irradiated nuclear materials has been explored. The Pb shielding was specially designed to reduce the detected radioactivity from the specimen during SANS experiments, and the overall configuration is termed shielded magnetic SANS (SM-SANS). Two FeCrAl-based alloys, C35M and 125YF, were examined with the SM-SANS technique using a free-form size distribution locally monodisperse model in both the as-received and irradiated states. Quantitative values derived from the free-form size distribution were compared with atom probe tomography experiments. Microstructural and compositional parameters determined using the two characterization techniques were complements of each other. The results demonstrate that the SM-SANS technique is an effective means of characterizing nanoscale clustering in irradiated material systems and provides new avenues for investigating radioactive material microstructures.

FeCrAl↗
DOE OSTI2020

Characterization of mechanical properties and deformation behavior of highly irradiated 316L stainless steel from target modules at the Spallation Neutron Source using digital image correlation analysis

Tensile specimens with irradiation doses ranging from approximately 0.7 to 7.6 displacements per atom (dpa) were fabricated from 316L samples removed from Spallation Neutron Source (SNS) target vessels after service. Tensile tests were performed in conjunction with digital image correlation (DIC) characterization, which permitted detailed analysis of the strain distribution and deformation behavior of the specimen gauge sections during testing. The results show, while considerable hardening (500–700 MPa increase in yield strength) occurred in the 316L SNS target material with increasing dose, appreciable ductility (19–23% total elongation) remained in the material after irradiation to 7.6 dpa in the mixed neutron/proton environment of the SNS target region. The DIC data showed the strain distribution and deformation behavior of all irradiated specimens were nonuniform during testing. Strain localization in the form of deformation bands or “hotspots” was observed during testing, and in some specimens, the bands propagated as deformation waves along the gauge section. The “traveling wave” deformation behavior affected the amount of elongation each specimen achieved, and the scatter in the elongation data was attributed to the large variation in deformation behavior observed among different specimens. Bands of localized strain, or deformation bands, oriented at specific angles were observed in two-dimensional plots of the strain distribution in the gauge sections of some specimens. Overall, this paper describes the experimental procedures, results, and deformation behavior of tensile specimens from SNS targets after service and illustrates the value of DIC characterization during tensile testing of irradiated materials.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Weldability of irradiated Stainless Steel 304 materials harvested from the National Research Universal (NRU) reactor

In this work, the weldability of the neutron-irradiated Stainless Steel 304 (SS 304) materials containing estimated helium levels of 12 to 44 appm has been studied through a Collaborative Research and Development Agreement between the Canadian Nuclear Laboratories and Oak Ridge National Laboratory. The work utilizes a SS 304 component harvested from the National Research Universal reactor. Laser beam welding was used to explore suitable welding conditions and associated parameters, to determine helium concentration limits for successful welding of irradiated stainless-steels. The experimental results show that maximum fusion zone void sizes, fusion zone void quantity and total heat affected zone helium-induced crack length generally increased with increasing He levels and effective weld heat input. A proper combination of welding parameters with low effective heat input, such as, laser power, weld travel speed and wire feed speed, were shown to improve the welding quality, and potentially reduce the length and probability of the formation of grain boundary helium-induced cracking and voids of irradiated SS 304 even with He concentrations as high as about 44 appm.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Defect Production and Microstructural Feature Impact for Radiation Damage in Additively Manufactured 316 Stainless Steel

This milestone presents multi-scale modeling research results for additively manufactured 316 stainless steel. A combination of phase field, cluster dynamics, molecular dynamics, and density functional theory with machine learning is used, allowing for predictions of radiation-driven microstructural evolution in additively manufactured 316 stainless steel over a range of temperatures, damage rates, neutron spectra, and microstructures, and supporting the development of combined ion and neutron irradiation for material qualification. Informed by ion irradiation and neutron irradiation results across the Advanced Materials and Manufacturing Technologies program, we investigate the unique aspects of radiation-driven microstructure evolution in additively manufactured 316 stainless steel. In particular, we focus on understanding the impact of carbon concentration (varying, for example, between the 316L and 316H standards) on void formation; radiation-induced segregation at dislocation cells and grain boundaries; and the evolution of dislocation loops and network dislocation populations. We find that the unique characteristics of the additively manufactured microstructures must be accounted for in understanding the evolution of dislocation populations under thermal and irradiation conditions, such as the variation in sink strengths arising due to the variation in dislocation density. We also find that the radiation-induced segregation of Cr and Ni to grain boundaries and cell walls differs due to the differences in their defect sink biases. We also find that increasing the Ni content can slow vacancy diffusion, which may provide a mechanism for the observed reduction in transient swelling rate for austenitic Fe-Cr-Ni alloys with increasing Ni content. In addition, ion irradiations have shown that increasing carbon content in 316 SS results in a larger population of smaller voids, suggesting reduced vacancy diffusion. Our results show that the carbon content of additively manufactured 316 SS has a significant impact on the migration rate of defect clusters. The presence of carbon atoms results in carbon-vacancy trapping, significantly reducing the diffusion rate of vacancies. Carbon atoms may also be trapped near the surface of a void, which may reduce void growth by trapping vacancies that diffuse toward the void.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗