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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Cavity Swelling in Irradiated Materials

Cavity swelling, due to three-dimensional clustering of vacancies, is an undesirable isotropic volume expansion of materials under irradiation. This phenomenon occurs above stage-III recovery temperature where vacancies are mobile, between ~0.3–0.6 Tm where Tm is the absolute melting temperature. The primary mechanism for cavity swelling is “dislocation bias”, i.e., preferential absorption of interstitials by network dislocations, causing a vacancy supersaturation. Cavity swelling is highly temperature dependent and typically follows a bell-shaped curve with peak swelling occurring at an intermediate temperature. With increasing temperature, cavity size increases, and number density decreases logarithmically. With increasing dose, cavity swelling increases, but follows a hockey-stick type pattern, characterized by a low-swelling transient incubation period and then a steady-state regime. The incubation period, dominant in cavity nucleation, is dependent upon material’s chemistry and irradiation conditions such as dose rate/temperature. Cavity swelling increases linearly with dose in the steady-state regime, that is independent of chemistry and dose rate. This regime is dominated by cavity growth. Transmutation gases like He enhance cavity nucleation such that an early onset of cavity formation occurs. However, swelling magnitudes can be either increased or suppressed when He is present. Results in austenitic steels and Cu indicate that cavity swelling peaks at intermediate He/dpa ratios of ~10 appm He/dpa. The most effective method to mitigate cavity swelling is by increasing the fixed point-defect sink density, by promoting precipitation or dispersion strengthening.

Bhattacharya, Arunodaya↗
DOE OSTI2020

Plasticity of irradiated materials at the nano and micro-scales

Here, we review here our recent work on plastic deformation in irradiated materials at the nano- and micro-scales, as revealed by Discrete Dislocation Dynamics (DDD) simulations. Two methods of including irradiation effects in the DDD framework are presented. The first directly captures the atomistic interaction mechanisms, while the second can effectively study high-dose irradiation. Computer simulations lead to new understanding of the dynamics of collective dislocation-irradiation defect interactions, as well as the quantitative analysis of the temporal and spatial characteristics associated with plastic instabilities. Based on these insights, theoretical models are developed to predict the critical conditions for dislocation channel formation. A simple probability model is proposed and demonstrated to predict the width of dislocation channels in bulk irradiated materials with good agreement with experimental data. The fundamental understanding of the origins of plastic flow localization in irradiated materials sheds light on the design of future generations of radiation-resistant materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the Advanced Test Reactor

The Advanced Test Reactor (ATR) is a light water reactor with aluminum-clad driver fuel. A primary mission of the ATR is to support the next generation of nuclear reactors. This support necessarily requires irradiation of advanced materials such as sodium, fuel salts, and metal eutectics. Irradiation of advanced materials in the ATR environment presents a challenge when completing accident analyses and demonstrating compliance to the Safety Analysis Report (SAR). Many advanced materials have the possibility to react with the ATR protective barriers such as the cladding or primary coolant system (PCS) boundary during postulated accident scenarios. Further, molten fuel experiments fall outside of the standard regulatory framework for dose consequence analyses. ATR is currently developing new safety analysis methods to support irradiation of advanced materials. The primary considerations for this development are 1) experiment containment design requirements, 2) primary coolant system response to an experiment containment failure, and 3) dose analyses for molten fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Safety Considerations for Advanced Material Irradiation at the ATR

The Advanced Test Reactor (ATR) is a light water reactor with aluminum-clad driver fuel. A primary mission of the ATR is to support the next generation of nuclear reactors. This support necessarily requires irradiation of advanced materials such as sodium, fuel salts, and metal eutectics. Irradiation of advanced materials in the ATR environment presents a challenge when completing accident analyses and demonstrating compliance to the Safety Analysis Report (SAR). Many advanced materials have the possibility to react with the ATR protective barriers such as the cladding or primary coolant system (PCS) boundary during postulated accident scenarios. Further, molten fuel experiments fall outside of the standard regulatory framework for dose consequence analyses. ATR is currently developing new safety analysis methods to support irradiation of advanced materials. The primary considerations for this development are 1) experiment containment design requirements, 2) primary coolant system response to an experiment containment failure, and 3) dose analyses for molten fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Generalized kinetic model for defect evolution in irradiated materials

A fundamental and open problem in materials science is to determine how the structural properties of irradiated and self-irradiated materials evolve in time. Structural defects generated by irradiation lead to changes in a material’s macroscopic properties, and these macroscopic changes, such as volumetric swelling, can in turn give rise significant alterations of the functionality of a material. Here, we develop a unified analytical model to understand the complex interplay between aggregation, fragmentation, and recombination processes involving mobile clusters of defects such as bubbles, voids, and interstitial atoms in materials subject to radiative sources. Specifically, we employ a mean field approach to derive a system of coupled kinetic equations that describes both the time evolution of the density of each type of defect and how clusters of each defect form and grow as the material ages.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Coupled cluster and dislocation dynamics modeling of microstructure evolution in irradiated materials

We develop here a coupled cluster and dislocation dynamics framework to study the microstructure evolution of irradiated materials. The framework not only accounts for the three dimensional diffusion of radiation-generated clusters, but also their interaction with dislocation networks and the resultant climb motion of discrete dislocations within finite crystals. The framework is solved with a superposition solution scheme, and is applied to investigate the evolution of the irradiation-induced dislocation loops in zirconium (Zr), considering the effects of various bias factors including the diffusion anisotropy difference (DAD) of interstitials and interstitial clusters, the dislocation bias of defects to discrete dislocation segments, and the production bias of defects from the radiation cascade. We find that the DAD is the most critical factor influencing the kinetics of the loop evolution in Zr, while the recombination/interaction of mobile defects can induce a strong spatial dependence of the loop evolution together with the DAD. Here, the method is also adopted to study the evolution of interstitial $\langle$a$\rangle$ and vacancy $\langle$c$\rangle$ dislocation loop ensembles consistent with the microstructure observed during irradiation-induced growth of Zr. Our findings not only reveal the spatial dependence of the size and ellipticity of the dislocation loops, but also suggest a limit on the anisotropy factor of interstitials to reproduce the co-growth of $\langle$a$\rangle$ and $\langle$c$\rangle$ loops in zirconium, in good agreement with experimental observations and other simulation results.

Bias factors↗
DOE OSTI2022

Optical Transmission Characterization of Fused Silica Materials Irradiated at the CERN Large Hadron Collider

The Target Absorbers for Neutrals (TANs) represent one of the most radioactive regions in the Large Hadron Collider (LHC). Seven 40 cm long fused silica rods with different dopant specifications, manufactured by Heraeus, were irradiated in one of the TANs located around the ATLAS experiment by the Beam RAte of Neutrals (BRAN) detector group. This campaign took place during the Run 2 p+p data taking, which occurred between 2016 and 2018. This paper reports a complete characterization of optical transmission per unit length of irradiated fused silica materials as a function of wavelength (240 nm - 1500 nm), dose (up to 18 MGy), and level of OH and H$_2$ dopants introduced in the manufacturing process. The dose delivered to the rods was estimated using Monte Carlo simulations performed by the CERN FLUKA team.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Light Water Reactor Sustainability Program: Complete the weld campaign on Ni-base irradiated materials using stress improved laser welding including the preliminary weld quality inspections

This report summarizes the most recent welding campaign on irradiated Ni-base alloy 182 and the preliminary weld quality inspections at the Radiochemical Engineering Development Center (REDC). Equipment and capabilities were developed jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program (and the Welding and Repair Technology Center), and Oak Ridge National Laboratory. Irradiated nickel alloy 182, with target helium contents of 5 atom parts-per million (appm), 10 appm, and 20 appm, were laser welded in the hot cell successfully. The significant, on-going effort to weld irradiated alloys with high helium concentrations and comprehensively analyze the results will eventually yield validated repair techniques and guidelines for use by the nuclear industry in extending the operational lifetimes of nuclear power plants. This report fulfills the FY 2023 milestone M3LW-22OR0406013, “Complete the weld campaign on Ni base irradiated materials using stress improved laser welding including the preliminary weld quality inspections”.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Conduct weld campaign (FY-21-1) on irradiated materials provided by the Canadian Nuclear Laboratory (CNL), including baseline post-weld evaluation and testing

A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. The objective is to explore suitable welding technique and parameters and to determine the helium concentration limitation in terms of irradiated stainless-steel weldability. An additional objective is to develop the knowledge and understanding of cracking mechanisms induced by high concentration of helium in the process of weld repair of the irradiated structural alloys. This report summarizes the experimental welding evaluation on irradiated material from the NRU reactor containing helium concentrations ranging from 12 to 45 atomic parts per million (appm).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Small scale tensile testing technique for measuring grain boundary strength of neutron-irradiated materials in focused ion beam systems

Irradiation induced defects on grain boundaries (GBs) can lead to GB strength degradation and cause intergranular cracking in structural materials, affecting the safe operation and service life of nuclear reactors. Establishing the quantitative relationship between the strength and character/microchemistry for individual GB in an effort of integrating experimental measurements and multiscale simulations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Challenges and successes in establishing a multipurpose fuels and materials irradiation capsule [Slides]

This presentation discusses the challenges and success related to designing and fabricating multipurpose capsules for irradiation experiments for irradiation within the Advanced Test Reactor (ATR) to be presented at tje DEVICE-MTR irradiation experiment design working group meeting hosted by the Japanese Atomic Energy Agency (JAEA). The information herein is limited to the design of the capsules only and no programmatic information from any project supporting the design of the multipurpose capsule is discussed. The intent of the design team was establish a capsule design that could accommodate multiple different irradiation project needs for the irradiation of nuclear fuel and materials of interest in the nuclear field. The main points of the presentation are as follows: use of tin as a molten thermal bond material was selected for the fueled capsule design and was found to be inadequate as it would corrode the stainless steel at high temperatures, this capsule design has high sensitivity to manufacturing of the gas gap and control of the gas mixture, and the fueled capsule has experienced welding challenges with the inner capsule material 800H Incoloy.

42 ENGINEERING↗
DOE OSTI2021

Positron annihilation spectroscopy of defects in nuclear and irradiated materials- a review

Positron is the only probe that can detect individual atomic vacancies and small and large vacancy clusters induced by irradiation with remarkable sensitivity, providing information about their size, concentration, and chemical environment. The focus of this review article is to provide guidance to facilitate applications of positron annihilation spectroscopy (PAS) in irradiation-induced defect studies to advance the development of new radiation-tolerant materials. The principle of PAS, its techniques, and data analysis methods are described. PAS studies of defects in nuclear and irradiated materials are reviewed and discussed in depth. Future developments to advance PAS applications in nuclear materials research and studies of materials under extreme environments are presented.

Atomic scale defects↗
DOE OSTI2021

Thermal Conductivity Measurement of Microstructure in Irradiated Materials

Argonne National Laboratory is establishing additional capabilities for advanced techniques for accelerated experiment-based fuel material screening. A key supporting element is expanded capabilities for the characterization of irradiated samples, including establishment of the capability to measure irradiated fuel thermal conductivity at the micro-scale level. Argonne is collaborating with the University of Illinois at Urbana-Champaign (UIUC) to adopt the suspended bridge method to measure thermal conductivity of radiation-induced microstructures in nuclear materials directly. The initial phase of work to support thermal conductivity measurements of microstructure in irradiated materials at Argonne has been executed, and is summarized here. This report is intended to serve as a status update for this activity, summarizing work completed in FY20.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Novel materials irradiation and pre-characterization plans for high-power targetry applications

The multi-megawatt proton beams that will be employed at future accelerator complexes introduce many new challenges for next-generation targetry systems, primarily due to the increased levels of beam-induced radiation damage combined with thermal shock effects during a beam pulse. Novel material classes, such as high-entropy alloys and electro-spun nanofibers are currently being investigated as materials that will be more tolerant of these effects in high beam power applications. Low-energy ion irradiation will be used for pre-screening and down selecting candidate novel materials prior to prototypic high-energy irradiation studies. Specific nano and micro-scale characterization techniques will be used to evaluate the radiation damage and thermal shock resistance of the new materials. Irradiation induced lattice spacing alteration and defect concentrations, as well as mechanical and thermal properties of the novel materials will be measured and compared to those of conventional targetry materials. This poster will detail plans for initial low-energy ion irradiation tests, and the post-irradiation examination techniques that will be used to develop and screen these novel target materials.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
NASA NTRS2019

Analyses and Methods of Solid Rocket Motor Material Irradiation at Marshall Space Flight Center

The search for life on other worlds is among humanity’s greatest endeavors. Europa represents the most probable location to discover extraterrestrial life in our solar system, owing to its surface composition of ice covering a liquid water ocean, warmed by the tidal forces of its orbit around Jupiter. Unfortunately, the Jovian system hosts the most intense planetary radiation environment in the solar system due to the charged particles, namely electrons and protons, trapped by Jupiter’s immense magnetic field. Any mission that attempts to approach or land on Europa must survive this radiation environment [Hand et al, 2017]. Radiation effects were identified as a priority risk to the successful development of a de-orbit stage and solid rocket motor (SRM) early in the Europa Lander De-orbit Stage project concept. The effects of primary concern tend to occur very near the outer surface of the SRM. The charged particles deposit their energy quickly and are mostly stopped in the outer metallic case, but a significant portion of radiation penetrates through the bondline and outer propellant regions. High doses of ionizing radiation are known to cause significant changes to mechanical properties of many materials, especially polymers. For polymers such as the rubber-like materials (elastomers) in a solid rocket motor, the primary damage mechanism is known as cross-linking, in which ionization causes the restructuring of the matrix of long polymer chains. Ionization energy breaks the long polymer chains and allows formation of new cross-linked bonding sites. This hardens and often strengthens the polymer, but at the cost of decreased flexibility (or modulus). Propellant, insulation, liner, and pyrotechnic materials were identified as higher risk items, and so were irradiated at Marshall Space Flight Center (MSFC) for investigation of changes in mechanical and ballistic properties. This process required significant levels of analysis to evaluate how the radiation environment evolves within the spacecraft during the mission, and also to evaluate how dose is delivered into test articles within the irradiating facilities.

Caffrey, Jarvis↗