Symposium Organizers
Maria Okuniewski, Purdue University
Chaitanya Deo, Georgia Institute of Technology
Maik Lang, University of Tennesee
Simon Middleburgh, Bangor University
EN17.01: Thorium Based Nuclear Fuels
Session Chairs
David Andersson
Maria Okuniewski
Monday PM, December 02, 2019
Sheraton, 3rd Floor, Hampton
8:30 AM - EN17.01.01
Fission Products in Thoria
Robin Grimes1,Navaratnarajah Kunganathan1,Partha Ghosh2,Ashok Arya2
Imperial College London1,Bhabha Atomic Research Centre2
Show AbstractWhile fuels based on thoria have been considered as alternatives for urania for decades, there is considerably less data concerning the accommodation of fission products at the atomic scale. Here atomic scale computer simulation based on density functional theory is used to predict the energies and accommodation sites for a range of fission products and He in stoichiometric ThO2 and hypo-stoichiometric ThO2-x.
Nine sites were considered including point defects and vacancy clusters with multiple species. Neutral and all possible defect charge states up to full formal charge were investigated. For Xe and Kr, in ThO2-x the most favourable solution equilibrium site is a neutral tri-vacancy while in ThO2 it is the di-vacancy. The most favourable solution site for I and Br is the single positively charged oxygen vacancy in ThO2-x while in ThO2, I demonstrates the same solubility in all clusters. Rb, Cs, Sr, Ba, Y and Zr are accommodated at the thorium vacancy but the charge state of the vacancy is important. Finally for He, a relationship is generated that describes the incorporation energy of the xth He atom, Ex (n, m), into a cluster consisting of n thorium vacancies and m oxygen vacancies.
9:00 AM - EN17.01.02
Radiation Effects in ThO2
Lingfeng He1,Tiankai Yao1,Vinay Chauhan2,Maniesha Kaur Salaken Singh3,Zilong Hua1,Marat Khafizov2,Anter El-Azab3,Matthew Mann4,Thierry Wiss5,Jian Gan1,David Hurley1
Idaho National Laboratory1,The Ohio State University2,Purdue University3,Air Force Research Laboratory4,European Commission Joint Research Centre5
Show AbstractOxide nuclear fuels have been widely used in light water reactors. The thermal conductivity of nuclear fuels is closely related to energy conversion efficiency as well as reactor safety margins. Understanding the mechanisms that cause the degradation of thermal conductivity in a high radiation environment is important for the design and development of new high-burnup fuels. For oxide nuclear fuels, phonon scattering by point defects, extended defects such as dislocation loops and bubbles, and grain boundaries plays a significant role in limiting the thermal transport properties. ThO2 is an actinide-bearing material that does not contain 5f electrons and will serve as a baseline that enables the investigation of phonon transport mechanisms in defective fuel without the effects of 5f electron. In this work, defect evolution in both single crystal and polycrystalline ThO2 has been studied by a combination of in situ ion irradiation and modeling. In addition, the effects of point defects and extended defects on the phonon transport in ThO2 have performed on ion irradiated samples. The microstructure of pristine and ion irradiated ThO2 has been characterized by using electron backscatter diffraction (EBSD), scanning transmission electron microscope (S/TEM), and time-domain Brillouin scattering (TDBS) techniques. The thermal conductivity before and after irradiation has been determined using laser-based modulated thermoreflectance (MTR) technique.
9:30 AM - EN17.01.03
Atypical Melting Behaviour of (Th,U)O2, (Th,Pu)O2 and (Pu,U)O2 Mixed Oxides
Robin Grimes2,Conor Galvin1,2,3,Patrick Burr1,Michael Cooper3,Paul Fossati4
University of New South Wales1,Imperial College London2,Los Alamos National Laboratory3,CEA4
Show AbstractNumerous experimental studies have investigated the melting point of mixed oxides in the ternary UO2–ThO2–PuO2 system, with significant discrepancies. In particular, a few studies have reported anomalous reduction in the melting point of mixed oxide, below that of the end members, which have variously been attributed to uncontrolled deviations of stoichiometry and limitations in the experimental setup. Here we provide a mechanistic understanding of the peculiar melting behaviour using molecular dynamics simulations, which are inherently free of the above limitations, combined with a robust pair-potential and a novel technique for calculating solidus and liquidus lines. We show that a dip in melting temperatures is indeed observed in the solidus and liquidus of the mixed oxides (Th,U)O2, (Th,Pu)O2 and (Pu,U)O2 and that this is an inherent property of the solid solutions. This dip is found at ∼5% additions of the oxide with higher melting point, in agreement with experimental observations. We propose that the root cause for this reduction in melting point is the formation of low-energy Frenkel pairs in the vicinity of small amounts of dissimilar cations of larger ionic radius: the random (and therefore locally heterogeneous) distribution of cations leads to a spread of Frenkel pair formation energies, depending on the local environment around the vacancy and interstitial. Overall this results in an increase in the average Frenkel pair formation energy, but a small portion of Frenkel configurations exhibit a significantly lower formation energy, which may act nucleation sites for early-onset melting.
9:45 AM - EN17.01.04
Equilibrium and Irradiation-Induced Point-Defect Disorder in ThO2 and U-Doped ThO2—Modeling and Ion Irradiation Experiments
Sanjoy Kumar Mazumder1,Maniesha Kaur Salaken Singh1,Tiankai Yao2,Lingfeng He2,Anter El-Azab1
Purdue University1,Idaho National Laboratory2
Show AbstractAs a part of EFRC Project titled ‘TETI: Thermal Energy Transport under Irradiation,’ the impact of irradiation induced defects in ThO2 and U-doped ThO2 is investigated. Prior to irradiation, the defect content of this material is tied to off-stoichiometry and hence it is dependent upon the U doping level and the temperature and external oxygen pressure or chemical potential. Under irradiation, non-equilibrium defects are produced by energetic particle bombardment, leading to the formation of nanoscale and sub-nanoscale defect clusters that influence thermal transport in ways that are not fully understood to the community. In this presentation we discuss our modeling results for the equilibrium and non-equilibrium defects in ThO2 and U-doped ThO2. A thermodynamic defect disorder model was used to investigate the off-stoichiometry behavior in ThO2 and U-doped ThO2 in an oxygen partial pressure and temperature environment. The model shows that while ThO2 remains mostly hypo stoichiometric at relevant thermodynamic conditions, U doping expands the thermodynamics window over which ThO2 becomes hyper-stoichiometric, thus illustrating the impact of 5f electrons introduced by the U doping on defect disorder. For example, U0.52Th0.48O2+x can sustain disorder up to an oxygen off-stoichiometry content of x = 0.4 at high temperatures, up to 2000 K, and high oxygen pressures, up to 0.316 atm, whereas U0.2Th0.8O2+x can only exist up to an oxygen off-stoichiometry content of x = 0.05. The extent of hyper-stoichiometry in the oxide reduces with increasing temperature and decreasing oxygen partial pressure values. We will also present the results of a recent cluster dynamics model of vacancy and interstitial cluster formation in ThO2 under irradiation in conjunction with microstructure and chemical analysis of pure and ion-irradiated ThO2 at different temperatures by EELS imaging technique in TEM. The impact of cluster composition on matrix composition is illustrated both computationally and experimentally, showing that irradiation provides a further mechanism of altering the stoichiometry of the matrix by forming clusters not commiserating with the chemical formula of the crystal.
EN17.02: Accident Tolerant and Advanced Nuclear Fuels
Session Chairs
Michael Cooper
Shenyang Hu
Monday PM, December 02, 2019
Sheraton, 3rd Floor, Hampton
10:30 AM - EN17.02.01
Atomistic Level Study of Oxidation of Ce3Si2 as an Accident Tolerant Nuclear Fuel Surrogate
Robert Harrison1,Robert Worth1,James Buckley1,Tim Abram1
University of Manchester1
Show AbstractSince the Fukushima accident in 2011 there has been a large international effort in the development of accident tolerant fuel and cladding materials for water cooled reactors [1]. Uranium silicide intermetallics are being considered as accident tolerant fuels (ATF) to replace UO2 currently used in light water reactors (LWRs) [2]. Primarily U3Si2 is being examined due to its good thermal conductivity (~15 W/m/K at 500°C) [3], higher U metal density, (giving economic advantages by requiring lower 235U enrichments) and reduced radiation swelling compared to the higher U containing silicides such as U3Si [4,5]. However, little is known on the oxidation behaviour of U3Si2 under high temperature air and steam environments which could be experienced under a loss of coolant accident (LOCA) scenario with water and/or air ingress into the fuel pin.
The oxidation process of Ce3Si2 (as a surrogate for U3Si2) has been studied to develop an oxidation mechanism. Samples were oxidised under flowing air to 750°C finding discrepancies between the terminal and theoretical mass gains for complete oxidation similar to previous works on the U3Si2 system. As oxidised materials were studied using scanning transmission electron microscopy (S/TEM) with energy dispersive x-ray (EDS) and energy filtered (EF)TEM mapping to elucidate the reaction products and develop a mechanism. The fate of the Ce and Si in the system will be presented and mechanistically discussed and discrepancy in the differences in mass gains revealed. The resulting composition and structure of the as-oxidised materials will also be presented, discussing potential consequences on mechanical stability of the fuel material under a LOCA scenario. The oxidation mechanism, comparison with the U-Si system and consequence of the resulting material structure on fuel performance under a LOCA will also be discussed.
[1] B.A. Pint, K.A. Terrani, Y. Yamamoto, L.L. Snead, Metall. Mater. Trans. E 2 (2015) 190–196.
[2] J.M. Harp, P.A. Lessing, R.E. Hoggan, J. Nucl. Mater. 466 (2015) 728–738.
[3] J.T. White, A.T. Nelson, J.T. Dunwoody, D.D. Byler, D.J. Safarik, K.J. McClellan, J. Nucl. Mater. 464 (2015) 275–280.
[4] I.J. Hastings, R.L. Stoute, J. Nucl. Mater. 37 (1970) 295–302.
[5] I.J. Hastings, J.R. MacEwan, L.R. Bourque, J. Am. Ceram. Soc. 55 (1972) 240–242.
10:45 AM - EN17.02.02
Specific Heat Measurements on USi from 2.4 K to 398 K
Jason Baker1,Joshua White1,Aiping Chen1,Robert Roback1,Hongwu Xu1
Los Alamos National Laboratory1
Show AbstractThe use of uranium-silicide compounds as potential nuclear reactor fuels requires knowledge of thermophysical properties as a function of temperature. At elevated temperatures, a great deal of research has been performed to understand the thermodynamic properties of U-Si compounds; however, in the low-temperature regime, only limited data are available [1-4]. Operating temperatures of nuclear reactors are high, yet important thermodynamic quantities such as Debye temperature and standard entropy (S°) can be derived from low-temperature measurements. S° is an imperative quantity in determination of the Gibbs free energy which depends on S° and enthalpy. Among the U-Si compounds, stoichiometric USi has typically received significantly less attention in the literature with regards to characterization; however, uranium-silicide phases used as nuclear fuels may include appreciable fractions of USi either introduced through fuel synthesis or fabrication. The potential of USi both as a nuclear fuel itself and as a fraction of other U-Si samples, implies a need to further the thermophysical understanding of this material.
Here we present investigations of the low-temperature thermal properties of USi by measuring its specific heat capacities from 2.4 to 398 K using a Quantum Design Physical Properties Measurement System (QD-PPMS). A smooth curve is observed over the entire temperature range measured, and the specific heat capacity at the maximum temperature tested (398 K) was determined to be 67.2 J mol-1 K-1. Additionally, the Debye temperature has been determined from fitting the low-temperature (below 30 K) specific heat capacity data as 252.3 K. Furthermore, by performing integration of the specific heat capacity over the temperature range, the standard entropy at room temperature has been determined as 88.5 J mol-1 K-1.
[1] J. T. White et al., J. Nucl. Mater., 2015, 456, 442-448.
[2] J.T. White et al., J. Nucl. Mater., 2016, 471, 129-135.
[3] J.T. White et al. J. Nucl. Mater., 2015, 464, 275-280.
[4] D.J. Antonio et al. J. Nucl. Mater. 2018, 508, 154-158.
11:00 AM - EN17.02.03
Fission Gas and Creep Behaviour in U3Si2 from DFT Calculations and Atomistic Simulations
David Andersson1,Michael Cooper1,Xiang-Yang Liu1,Benjamin Beeler2,Kyle Gamble2,Giovanni Pastore2
Los Alamos National Laboratory1,Idaho National Laboratory2
Show AbstractThe U3Si2 compound has a high uranium density and thermal conductivity compared to standard UO2 fuel and, for these reasons, it is being considered as a possible replacement fuel in the current fleet of light water nuclear reactors. It could bring significant performance and economic benefits, while the exothermic reaction with the coolant upon cladding breach is a known drawback that needs to be addressed. Material properties such as the diffusion rate of fission gas atoms and point defects must be determined in order to model the in-reactor behaviour of U3Si2. Here we use density functional theory (DFT) calculations and empirical potential simulations to study diffusion of Xe and point defects in bulk U3Si2 as well as at grain boundaries. The DFT calculations apply the GGA+U methodology to calculate defect energies, entropies and migration barriers of point defects as well as of Xe atoms interacting with point defects. Xe and point defect diffusion rates are predicted from a point defect model parameterized by results from DFT calculations. The diffusion rate under irradiation due to ballistic damage is derived from molecular dynamics simulations based on a MEAM potential for the U-Si-Xe interactions. Molecular dynamics simulations are also used for investigating grain boundary diffusion and segregation of point defects and Xe atoms. We find that Xe diffusion in U3Si2 is faster than in conventional UO2 nuclear fuel in the intrinsic regime, but lower in the regime where transport is dominated by ballistic mixing. The latter observation is a consequence of the high thermal conductivity of U3Si2 preventing high-temperature thermal spikes, which leads to transport by ballistic mechanisms being dominant. The impact of the diffusion properties on fission gas release and swelling in U3Si2 is investigated by parametrizing a fuel performance model in the Bison code with results from the atomic scale simulations. The results are compared to recent irradiation experiments available in the literature. Finally, we discuss the importance of bulk and grain boundary diffusion on creep rates in U3S2.
11:15 AM - EN17.02.04
Ceramic Oxide Coatings for Accident Tolerant Fuel Concept in Light Water Reactors
Fabio Di Fonzo1,Mattia Cabrioli1,2,Matteo Vanazzi1,Erkka Frankberg1,Koba Van Loo3,Jozef Vleugels3,Konstantina Lambrinou4
Istituto Italiano di Tecnologia1,Politecnico di Milano2,KU Leuven3,SCK-CEN4
Show AbstractAccident Tolerant Fuel (ATF) concepts aim at providing radical performance and safety improvements for the Light Water Reactors (LWR) at economically attractive conditions. Coatings as near-term evolutionary option take advantage of the reliability of traditional Zirconium-based cladding materials for structural purposes, while conferring engineering of the surface properties. In particular, they are expected to reduce corrosion rates, retain structural integrity and adhesion, minimize hydrogen uptake by the cladding matrix and improve high temperature oxidation resistance.
In this work, the process of selection, design, production and testing of ceramic oxides as protective coatings for LWR fuel cladding is presented. It is worth mentioning that all these activities have been performed in the framework of the European project IL TROVATORE. In the first stage, candidate materials are selected by matching neutronic requirements and chemical stability in pressurized-water environment. This phase is followed by the production of ceramic oxides, in the form of sintered pellets or as thin films by the Pulsed Laser Deposition (PLD) technique. Specimens are then exposed to Pressurized Water Reactor (PWR) environment. Specifically, compatibility tests are performed in pure water at 360°C and saturation pressure, in order to simulate the PWR core during normal operating conditions. Mass changes are monitored during the exposure, then correlated to modifications on the surface and in the structure of the tested samples. The specific dissolution and oxidation processes related to the measured mass changes are also investigated. This initial phase allows us to simultaneously evaluate the behaviour of candidate materials as bulk oxides and coatings, hence confirming the feasibility of PLD as a viable technique to produce high-performance ceramic films, with strong adhesion and interfacial bonding. Eventually the screening phase leads us to the selection of best candidate oxide to uptake an additional phase of optimization of the coating fabrication process.
The selected material is grown on relevant substrates such as Zirconium-based alloys and AISI 316L via an optimized PLD process. The exposure of the new coated specimens to the simulated PWR ambient is performed, up to 30 days in saturated pure water at 360°C. The analysis of the mass change confirms that the coatings act efficiently as barriers against oxidation by reducing significantly the oxidation of the metal alloy. In addition, structural integrity and good adhesion to the substrates are confirmed by spectroscopic characterization and electron microscopy.
Furthermore, the improvement of tolerance against severe accidents is considered with respect to the specific case of Loss Of Coolant Accident (LOCA). Preliminary tests at high temperature (up to 1000°C) and in a controlled atmosphere are conducted in order to evaluate the performance of the selected ceramic material during accidental transients. The coatings show structural integrity and good adhesion, suggesting the high capability of protecting the underlying alloy and their potential for a relevant delay of uncontrolled oxidation of the claddings during accidental transients. On the other hand, the importance of correct alignment of the thermal expansion coefficients between coating and substrates becomes evident at higher temperatures (≥1200°C).
In conclusion, the overall results collected show that PLD ceramic coatings are capable of sustaining both normal and off-normal reactor conditions and consequently constitute a promising option for coated-cladding ATF concepts.
11:30 AM - EN17.02.05
DFT+U Point Defect Calculations of Uranium Mononitride Ground State and Metastable States
Bryant Jerome1,Dilpuneet Aidhy1
University of Wyoming1
Show AbstractUranium mononitride (UN) is a proposed nuclear fuel for upcoming generation IV nuclear reactors, due to its higher thermal conductivity, greater melting point, and higher concentration of fissile material compared to conventional oxide fuels DFT+U is used to capture the anti-ferromagnetic ground state of UN. However, the +U correction leads to the possibility of convergence to metastable states, thereby disrupting the correct predictions of electronic-level properties, particularly point-defect energies. In this work, we provide a comprehensive analysis of ground and metastable states and the resulting defect properties using three methods, namely occupation matrix control (OMC), quasi-annealing (QA), and U-ramping. We preform each of these calculations using both the Dudarev and Liechtenstein rotationally-invariant forms of DFT+U. We find that QA and OMC reproduce the ground state, whereas U-ramping converges to metastable states as previously observed. All three methods predict the nitrogen interstitial to be the most favorable defect, whereas the uranium interstitial is the least favorable defect among the vacancies and interstitials of both elements. For example, the calculated formation energies of defects calculated using the QA method across the Dudarev form are 1.76 eV and 9.38 eV for the U vacancy and interstitial, respectively, and for the N vacancy and interstitial are 3.86 eV and 0.94 eV , respectively. The migration barriers for both the U and N vacancies are similar, i.e., ~ 3 eV. Finally, our results show there are significant deviations in defect formation energies among various metastable states.
11:45 AM - EN17.02.06
Multifunctional Nano-Ceramic Coatings—The Enabling Technology for Next Generation Nuclear Reactors (Including Fusion)
Fabio Di Fonzo1,Matteo Vanazzi1,Mattia Cabrioli1,Boris Paladino1,Daniele Iadicicco1,Erkka Frankberg1
Istituto Italiano di Tecnologia1
Show AbstractNext generation nuclear systems are meant to outperform current ones, by providing disruptive solutions in terms of non-proliferation, fuel cycle efficiency, radioactive waste management, safety and economics. However, the real occurrence of this scenario is directly linked to the availability of suitable materials for the more demanding conditions of these advanced plant concept, in terms of higher burnup and operating temperature, intense radiation fields, liquid metal or molten salt corrosion and tritium permeation. Interposing an engineered surface layer between qualified structural materials and the reactor environment makes this scenario reachable in the short term. In the last years our group developed multifunctional nanoceramic coatings by Pulsed Laser Deposition and Atomic Layer Deposition for Accident Tolerant Fuel for Light Water Reactors and as a solution to high-temperature Heavy Liquid Metal (HLM) corrosion in Generation-IV (GIV) and fusion systems. Accident Tolerant Fuel (ATF) concepts are developed by depositing on ZIRLO® and Zircaloy-4 suitable nanoceramic coatings, which are found to minimize hydrogen uptake and to improve high temperature oxidation resistance. On the other hand, in respect to GIV and fusion systems, Al2O3 and Y2O3 on AISI316, 1515-Ti and EUROFER-97, has been tested as anti-corrosion, radiation-resistant tritium permeation, insulating barriers. In particular, the compatibility in Pb and Pb-Li has been proven up to 10,000 hours. The tritium permeation reduction of these films is in the order of 104-105, well above the design requirements. Furthermore, Al2O3 coatings have been tested under heavy ions irradiation, at damage levels relevant for fission and fusion applications. The ceramic film under irradiation preserves its integrity and mechanical behaviour, evolving structurally from an amorphous to a crystalline state. To conclude, engineered coatings deposited by PLD/ALD techniques represent promising candidates to face the major issues related to future nuclear technologies and allow the design of innovative and economically attractive power plants.
EN17.03: Metallic Nuclear Fuels
Session Chairs
Gianguido Baldinozzi
Maik Lang
Monday PM, December 02, 2019
Sheraton, 3rd Floor, Hampton
1:30 PM - EN17.03.01
Microstructure-Based Model of Gas Bubble Swelling in Polycrystalline UMo Fuels by Integrating Cluster Dynamics and Phase-Field Approaches
Shenyang Hu1,Benjamin Beeler2,Douglas Burkes1
Pacific Northwest National Laboratory1,Idaho National Laboratory2
Show AbstractExperiments show that recrystallization dramatically speeds up the gas bubble swelling kinetics in UMo nuclear fuels. This implies that gas bubbles inside the recrystallization zone, which has hundred nano-meter sized grains in diameter, grow much faster than those inside coarse grains. However, the mechanism of fast gas bubble growth is not well understood. In this work, a gas bubble evolution model integrating cluster dynamics and phase field models is developed to study the effect of vacancy and interstitial clustering on gas bubble evolution and volumetric swelling kinetics in the recrystallization zone. Cluster dynamics is used to describe the evolution of interstitial loops, vacancy clusters, and dislocation density. The phase field model is used to describe the gas bubble nucleation and growth. With this integrated model, the effect of defect generation rate, defect clustering rate, defect sink strength, and interstitial emission rate from interfaces on gas bubble evolution were systematically simulated. The comparison of gas bubble size, density and volumetric swelling from simulations and experimental data demonstrate that the nucleation and growth of interstitial loops in recrystallization zone is one of key mechanisms behind the fast swelling kinetics.
2:00 PM - EN17.03.02
Modeling Irradiation Induced Grain Refinement Utilizing Cahn’s Time Cone Method
Alejandro Figueroa1,Joshua Pribe1,Walter Williams1,2,Rayaprolu Goutham Sreekar Annadanam1,Edwin Garcia1,Maria Okuniewski1
Purdue University1,Idaho National Laboratory2
Show AbstractIrradiation induced grain refinement is a high burnup phenomenon associated with the fuel’s response to high strain fields created through in reactor irradiation. The increase in grain boundary area from the refinement increases the nucleation rate of fission gas bubbles which leads to a sharp increase in the swelling rate of the fuel. Understanding this phenomenon will allow for more accurate fuel life predictions and can lead to more efficient fuel design. The fission process creates multi-dimensional defects, including fission products, which strain the lattice. The defects produced by this process propagate through material creating both pinning points for dislocations and grain-boundaries, as well causing dislocation propagation by acting both as a source and a climb mechanism. This increases the internal energy of the material, which provides a driving force for recrystallization. Current models of this phenomenon approach the behavior sequentially; addressing grain nucleation first, then growth of the nuclei. This approach is useful but does not fully capture the temporal inhomogeneity of the system. By utilizing Cahn’s time cone method and adapting it to a temporally inhomogeneous reactor environment the nucleation and growth of the irradiation induced grain refinement can be modeled concurrently and can provide an analytical solution to the phenomenon. To address the temporal inhomogeneity of the system, strain which affects both the nucleation and growth was defined as a function of fission density. The mobility of the system was also adapted for irradiation enhanced diffusion. Heterogenous nucleation was used as the dominant nucleation mechanism. Utilizing this technique to model irradiation induced grain refinement showed comparable results to experimental observations demonstrating that the adapted Cahn time cone method can effectively model the phenomenon, while addressing both nucleation and growth simultaneously.
2:15 PM - EN17.03.03
The Effects of Fabrication Parameters on the Microstructure of Monolithic U-Mo Nuclear Fuels
Jan-Fong Jue1,Dennis Keiser1,Adam Robinson1,Brandon Miller1,Jian Gan1,Glenn Moore1
Idaho National Laboratory1
Show AbstractMonolithic U-Mo fuels have been down-selected for the conversion of several US high performance research and test reactors. In order to qualify this new fuel type, more than a hundred fuel plates have been fabricated and irradiated to date in the Advanced Test Reactor at Idaho National Laboratory. The size of fuel plates fabricated for these irradiation campaigns varies from 4 inches (mini-plates) to more than 40 inches (full-size plates) in length. The fabrication parameters used in fabricating mini-plates and full-size plates are very different, thus resulting in different as-fabricated microstructures. The microstructure of as-fabricated and irradiated mini-plates and full-size plates will be presented. The impact of the fabrication parameters on the microstructure and irradiation performance will be discussed.
2:30 PM - EN17.03.04
Density Functional Theory Study of Uranium-Based Compounds
Edmanuel Torres1,Thaneshwor Kaloni1,Jeremy Pencer1
Canadian Nuclear Laboratories1
Show AbstractThe accurate description of the electronic and thermal properties of uranium-based compounds is fundamental in the development of novel nuclear fuels. Density functional theory calculations (DFT), within the plane wave pseudopotential approach using Quantum ESPRESSO (QE), has been applied to the study of uranium-based compounds. The lack of widely accessible pseudopotentials to perform DFT studies of uranium compounds significantly limits the capability of innovation in nuclear materials. As such, we have developed a set of projector augmented-wave (PAW) pseudopotentials for uranium, oxygen, nitrogen, and carbon atoms. The pseudopotentials were optimized to be used with a low kinetic energy cutoff of 37 Ry, which is considerably lower in comparison to PAW pseudopotentials available for QE. The set of pseudopotentials is then used to perform DFT calculations of uranium-based compounds, such as uranium metal, uranium oxides (UO2, UO3, and U3O8), uranium mononitride, and uranium monocarbide. Systems with strong electron correlations were described using Hubbard corrected DFT+U calculations. As such, monitoring the occupation matrix for 5f electrons of the uranium atoms was required to avoid metastable state solutions, and therefore to obtain reasonable ground properties. The reported results are shown to be accurate in comparison to reported theoretical and experimental results. This new set of PAW pseudopotentials are thus suitable for use in modelling a variety of uranium compounds, but also have the advantage of enhanced computational efficiency, due to the reduced kinetic energy cutoff.
2:45 PM - EN17.03.05
Defect Clustering in Irradiated Alpha Uranium—Cluster Dynamics Modeling and Ion Irradiation Experiments
Sanjoy Kumar Mazumder1,Fabia Farlin Athena1,Tiankai Yao2,Lingfeng He2,Anter El-Azab1
Purdue University1,Idaho National Laboratory2
Show AbstractThere is limited information in the literature about defect clustering in irradiated alpha uranium in spite of the fact that defect clustering is responsible for irradiation growth of that material. The limited available data shows that two types of dislocation loops form in alpha uranium on the (010) and (100) planes, with Burgers vectors of $\sqrt{a^2+b^2} <110>/2$ and $a [100] $, respectively, under neutron irradiation. Under He ion irradiation, however, small cavities, which are assumed to be He bubbles were observed. We present a cluster dynamics model for loop formation in alpha uranium. The defect clusters considered at prismatic loops of interstitial type, those on (010) planes, and vacancy type, those on (100) planes. The dynamics model parameters were partly taken from open literature and partly fixed by molecular dynamics simulations. Cluster dynamics simulations were carried out and at different temperatures and dose rates, and the comparison with the existing neutron irradiation data for the loop size distribution and average loop size and density shows that the model is able to make reasonable predictions. We also present the results of proton irradiation of alpha uranium, where the proton beam was used to produce a microstructure dominated by dislocation loops. The formation and evolution of dislocation loops were studied by TEM technique, which provided further data for validation of the cluster dynamics model.
EN17.04: Oxide Nuclear Fuels
Session Chairs
Lingfeng He
Simon Middleburgh
Monday PM, December 02, 2019
Sheraton, 3rd Floor, Hampton
3:30 PM - EN17.04.01
Structural and Microstructural Features Observed in (Ln,U)O2-x Systems
Gianguido Baldinozzi1,Bernardo Herrero-Bocco1,2,Fabienne Audubert2,Lionel Desgranges2,Luis Casillas3,Maulik Patel4,Haixuan Xu5,Kurt Sickafus5
University of Paris Saclay1,CEA2,Linköping University3,University of Liverpool4,The University of Tennessee, Knoxville5
Show AbstractThe majority of periodic table elements are created within the conventional uranium dioxide fuel matrix during its time in pile. These elements are accommodated in quite different ways within the nuclear ceramic. Ba, Zr, Y and the lanthanide atoms are typically substituted within the uranium dioxide matrix. At high burnup, the number of phases that form and the differences between them can result in the development of highly complex structures and microstructures. Similar effects might be encountered in manufactured advanced actinide fuels for the transmutation of minor actinides or in fuels containing significant amounts of burnable poisons (typically Gd). Those systems constitute then a large class of compounds, generally with fluorite or fluorite-derived structures, exhibiting high radiation tolerance, possibly related to their peculiar ability to accommodate a variety of defects and to exist as nonstoichiometric compounds within a large homogeneity range. Nevertheless, a variety of behaviours and different phases is observed after quenching or annealing in several of the peudo-binary systems of type (Ln,U)O2-x, suggesting that the thermal history of those systems is of paramount importance. Some of these systems, like (U,Ce)O2-x are particularly complex, display an intriguing structural variety, and therefore present a fundamental interest. In this talk we would try to address some of those interesting structural features induced by aliovalent substitutions in UO2.
4:00 PM - EN17.04.02
Modeling Fission Gas Behavior in Doped UO2\
Michael Cooper1,Topher Matthews1,Kyle Gamble2,Giovanni Pastore2,Chris Stanek1,David Andersson1
Los Alamos National Laboratory1,INL2
Show AbstractAdvanced fuels are being considered for deployment in light water reactors (LWRs). Advanced fuels are designed to improve performance and fuel economics during normal operation and/or safety margins during accidents. For example, additives such as Cr2O3 can be used during fuel fabrication to modify the UO2 microstructure. Specifically, the resultant enhanced grain size is considered to benefit pellet-clad interactions by modifying the mechanical properties of the fuel and by increasing fission gas retention, which reduces the plenum pressure. The effect of increased grain size on fission gas release can be readily captured within physics-based fission gas models, such as that implemented in the BISON fuel performance code. However, the dopant could modify the defect chemistry of UO2 during reactor operation, thus altering the fission gas diffusivity and potentially undermining the enhanced fission gas retention obtain through enlarged grains. Fission gas diffusivity exhibits three diffusion regimes for reactor conditions: high T intrinsic diffusion, intermediate T irradiation enhanced diffusion, and low T athermal diffusion. In this work, we present a cluster dynamics model developed and validated for the description of irradiation enhanced diffusion in undoped UO2. Large defect clusters play a crucial role in the reproduction of irradiation enhanced diffusion. For doped UO2, it is assumed that the oxygen potential is controlled by the preferential reduction of Cr2O3 to Cr (over UO2+x reduction to stoichiometric UO2). The resultant doped UO2 Xe diffusivity is enhanced with respect to undoped UO2. The enhancement in Xe diffusivity relative to undoped UO2 has been implemented in BISON. The predicted fission gas release from BISON has been compared to Halden experiments.
4:15 PM - EN17.04.03
Prediction of the Shape, Size and Pressure of Intragranular Fission Product Bubbles in Spent Nuclear Fuel
Michael Rushton1,Conor Galvin2,3,Michael Cooper4,Patrick Burr2,Simon Middleburgh1,William Lee1,3,Robin Grimes3
Bangor University1,University of New South Wales2,Imperial College London3,Los Alamos National Laboratory4
Show AbstractDuring its life in a reactor fission products build up in nuclear fuel. Some of these such as Xe, Kr and He form bubbles in the fuel. Whilst in a reactor, these have a detrimental effect on fuel performance as fission product bubbles may lead to swelling and a degradation of mechanical properties, and when released from the fuel, where the gas degrades the thermal conductance of the fuel-clad gap and leads to an increase in fuel temperature and cladding pressure. Understanding the size, shape and pressure of bubbles also allows the fission product inventory of nuclear fuel to be gauged. This information is important in developing improved fuel performance codes, pushing fuel to higher burn-up and has relevance to back-end processes such as reprocessing and ultimate disposal of spent fuel.
In this work, molecular dynamics simulations have been used to study fission product bubbles in UO2. The results of this work will show the behaviour of small fission product bubbles in the range of diameters below <100nm. In particular bubble morphology is considered with both spherical and faceted Wulff shapes being compared. The effect of bubble internal pressure on morphology and its dependence on bubble size will be presented. The simulation results will be described in terms of the bubble size and pressure distribution expected in spent fuel.
4:30 PM - EN17.04.04
Effect of High Burnup Structure on Fuel Rod Performance with Burnup Extension
Hongbin Zhang1,Jianguo Yu1
Idaho National Laboratory1
Show AbstractThe formation of high burnup structure (HBS) or rim structure is possibly the most significant restructuring processes at the rim of pellets in-pile with burnup extension in LWR and the effect of HBS on fuel thermo-physical or mechanical properties is a key requirement to ensure the successful implementation of extended burnup. There is independently experimental evidence to support that HBS concomitantly affects thermal and mechanical properties of fuel. But it is still not well understood how such HBS affects fuel thermo-physical or mechanical properties. In this work, we present results how HBS impacts the fission gas behavior of fuel rod under normal operating conditions. The fuel performance code FRAPCON-4.0 was used to simulate fission gas related properties, such as plenum pressure, fission gas release, and gap conductance. The effect of HBS under different boundary conditions, i.e. power history profiles, will also be discussed.
4:45 PM - EN17.04.05
Fracture Behavior of Pure UO2 Using Small-Scale Mechanical Testing
Brent Heuser1,Shen Dillon1
University of Illinois at Urbana Champaign1
Show AbstractFracture of UO2 affects fission gas release in fuels. Predicting, understanding, and controlling within UO2 fuel pellets is of considerable practical interest. An ongoing fundamental challenge is understanding the constituent contributions form grain boundaries and the bulk to the fracture response of the polycrystal. In this work, micro-scale mechanical testing has been used to induce single grain transgranular fracture and bicrystal intergranular fracture in samples isolated from polycrystalline UO2. The results are discussed in the context of bulk UO2 fracture and similar measurements performed on other oxides.
Symposium Organizers
Maria Okuniewski, Purdue University
Chaitanya Deo, Georgia Institute of Technology
Maik Lang, University of Tennesee
Simon Middleburgh, Bangor University
EN17.05: Fusion Materials
Session Chairs
Tuesday AM, December 03, 2019
Sheraton, 3rd Floor, Hampton
8:30 AM - EN17.05.01
Damage Induced in Self Irradiated Tungsten—Effect of Irradiation Conditions and Purity
Marie-France Barthe1,Zhiwei Hu1,Cécile Genevois1,Brigitte Decamps2,Pierre Desgardin1,Robin Schaüblin3
CEMHTI CNRS1,CSNSM CNRS2,ETHZ3
Show AbstractTungsten has been chosen to cover the divertor and is envisaged for first wall material in fusion power reactors because of its high melting point, good thermal conductivity, low thermal expansion, high strength at high temperatures and high sputtering threshold energy. In such systems, tungsten will be submitted to neutron irradiation, high Helium and Hydrogen fluxes, and high heat fluxes up to 10 MW.m-2 in stationary and up to 20 MW.m-2 in transient operation and will have to sustain a temperature up to 1780 K. Such severe operating conditions could have a high impact on the macroscopic properties of the material, such as embrittlement and swelling.
In order to study and understand the evolution of the microstructure of tungsten in future fusion reactors such as ITER and DEMO, well-prepared tungsten samples were irradiated with W ions at 1.2 MeV and 20 MeV. Different damage doses in the range from 0.01 to 0.1 dpa are investigated and irradiation temperatures are fixed in the range from RT to 700°C. Positron annihilation spectroscopy (PAS) and Transmission Electron Microscopy (TEM) are used to characterize defects induced by irradiation. It allows to detect from the small vacancy clusters to large cavities and, for some conditions, the dislocation loops are also studied. The defects size, their number density and for the latter the nature of their Burgers vectors is determined.
The defects distribution is investigated as a function of the irradiation conditions and of the purity of the samples. Their evolution is then followed as a function of the damage dose, the irradiation temperature and also of the annealing temperature after irradiation. The results will be discussed in regard of the role of impurities on the microstructure evolution in the first stages of irradiation in tokamak and of the impact on the swelling of the material.
Acknowledgements:
This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Irradiations were performed at JANNuS (Joint Accelerators for Nanoscience and Nuclear Simulation) Orsay at CSNSM (France) and Saclay at CEA (France), which are part of the EMIR French accelerators network.
8:45 AM - EN17.05.02
Void Evolution in Tungsten and Tungsten-5wt.% Tantalum under In Situ Proton Irradiation
Robert Harrison2,Iuliia Ipatova1,Simon Middleburgh1,Michael Rushton1,Enrique Jimanez-Melero2,Stephen Donnelly3
Bangor University1,University of Manchester2,University of Huddersfield3
Show AbstractVoid evolution during in-situ irradiation in polycrystalline W and W-5wt.%Ta material has been examined at 800 and 1000°C, via TEM. No dislocation loop formation was observed prior to void nucleation. The irradiated W microstructure was characterised by the presence of a population of voids, randomly distributed in the region. The number density of the voids reduces when the irradiation temperature is raised. In contrast, the excess of free vacancies in the W-5wt.%Ta material irradiated at 800°C only leads to the formation of visible voids after post-irradiation annealing of the sample at 1000°C. Solute Ta atoms also cause a significant decrease in the number density of voids when comparing the microstructure of both materials irradiated at 1000°C, and a gradual progression towards saturation at 0.2 dpa. Moreover, for the first time, we have detected a progressive transition from a spherical to a faceted shape in a number of voids present in both materials at damage levels 0.3 dpa.
9:00 AM - EN17.05.03
High Irradiation Resistance of Nanocrystalline W-Based High Entropy Alloy
Enrique Martinez1,Osman El Atwani1,Nan Li1,Meimei Li2,Arun Devaraj3,Kevin Baldwin1,Matthew Schneider1,Damian Sobieraj4,Jan Wrobel4,Duc Nguyen-Manh5,Stuart Maloy1
Los Alamos National Laboratory1,Argonne National Laboratory2,Pacific Northwest National Laboratory3,Warsaw University of Technology4,Culham Center for Fusion Energy5
Show AbstractA novel W-based refractory high entropy alloy with outstanding radiation resistance has been developed. The alloy was grown as thin films showing a bimodal grain size distribution in the nanocrystalline and ultrafine regimes and a unique 4 nm lamella-like structure revealed by atom probe tomography (APT). Transmission electron microscopy (TEM) and X-ray diffraction show an underlying body-centered cubic crystalline structure with certain black spots appearing after thermal annealing at elevated temperatures. Thorough analysis based on TEM and APT correlated the black spots with second phase particles rich in Cr and V. After both in situ and ex situ irradiation, these precipitates evolve to quasi-spherical particles with no sign of irradiation-created dislocation loops even after 8 dpa at either room temperature or 1073 K. Furthermore, nanomechanical testing shows a large hardness of 14 GPa in the as-deposited samples, with a slight increase after thermal annealing and almost negligible irradiation hardening. Theoretical modeling based on ab initio methodologies combined with Monte Carlo techniques predicts the formation of Cr and V rich second phase particles and points at equal mobilities of point defects as the origin of the exceptional radiation tolerance. The fact that these alloys are suitable for bulk production coupled with the exceptional radiation and mechanical properties makes them ideal structural materials for applications requiring extreme conditions.
9:15 AM - EN17.05.04
Modified Deformation Behavior of Ion-Implanted Tungsten
Felix Hofmann1,Suchandrima Das1,Hongbing Yu1,Edmund Tarleton1
University of Oxford1
Show AbstractTungsten is the main candidate material for plasma-facingarmor components in future fusion reactors. During operation bombardment with high-energy fusion neutrons will create collision cascades that leave behind lattice-defects. Helium, injected from the plasma and produced bytransmutation, strongly binds to these defects, and modifies their behavior and retention. We investigate the impact ofhelium-implantation-induced and self-ion-implantation-induced damage on the deformation behavior of tungsten, comparing spherical nano-indents in unimplanted, helium-implanted and self-ion-implanted tungsten crystals of 001-orientation. Ion-implantation increases hardness and causeslarge pile-ups that increase with increasing implantation dose. Lattice rotations and indentation-induced residual strains beneath indents, probed using 3D-resolved synchrotron X-ray micro-diffraction, are smaller in the ion-implantedmaterial, suggesting a more confined plastic zone. The increase in pile-up points to a reduction in strain hardening capacity or even a strain softening. Based on these results, as well as TEM observations of defects beneath indents, we hypothesize that while dislocationmotionis initially obstructed by implantation-induced defects, the strength ofthese obstacles is reduced by the passage of dislocations. A 3D crystalplasticity finite element (CPFE) model founded on this hypothesis, with only three fitting parameters, reproduces these effects and is directly compared to the experiments. We also explore the orientation dependence of the indentation-response by considering grains with <001>, <110> and <111> out-of-plane orientations. Indents in unimplanted tungsten show little orientation dependence. In the helium-ion-implanted material, <110> and <111>-oriented grains show a much lower pile-up and hardness than <001>-oriented grains. The CPFE formulation, with all parameters unchanged except the crystal orientation, captures these orientation-dependent changes. The results suggest an orientation-independent mechanism governing the interaction of implantation-induced defects with glide dislocations, with differences in pile-up morphology arising just due to the relative orientations of the crystal slip systems, sample surface and spherical indenter tip.
9:30 AM - EN17.05.05
Microstructural Features and Hydrogen Isotope Diffusion and Retention in Ion Irradiated Lithium Aluminate
Weilin Jiang1,David Senor1
Pacific Northwest National Lab1
Show AbstractAs a tritium (T) production material, 6Li enriched tetragonal lithium aluminate (γ-LiAlO2) has been used in support of the Tritium Sustainment Program. It also has been investigated as a candidate blanket material for fusion reactor designs. When γ-LiAlO2 is irradiated with thermal neutrons, 2.75 MeV T and 2.05 MeV He particles from reaction 6Li (n, He) T will emit. These energetic particles collide with the lattice atoms and initiate damage cascades, generating point defects in the crystalline structure. The accumulation and interaction of point defects and gas species lead to the formation of extended defects and bubbles. In addition, radiolysis due to electronic energy deposition causes the material to decompose. The modified microstructure will affect T diffusion in the irradiated material.
In order to gain a physical insight into the defect production and gas species diffusion processes, ion irradiation experiments have been designed and performed in reactor-relevant conditions. Both protium and deuterium (D) have been used as surrogates for T in this study. Sequential irradiation with 90 keV He+ and 80 keV H2+ ions as well as 120 keV He+ and 80 keV D2+ ions with an equal ion fluence was performed over temperatures from 188 to 773 K. The ion energies were so chosen that the H or D profile peak was located at the He+ ion damage peak. The ion fluence ranged from 2×1016 to 4×1017 He++H+ (or He++D+)/cm2. Isochronal and isothermal annealing experiments were followed up to 873 K and 60 min, respectively. The irradiated samples were characterized using a number of spectroscopy and microscopy techniques, including RBS/C, ToF-SIMS, STEM, STEM-EDS, STEM-EELS, HIM, Nano-SIMS and APT.
This presentation will report our main results from ion irradiation studies of monocrystalline γ-LiAlO2 and polycrystalline pellets. Major conclusions from these studies include the following. Saturation stages of lattice disorder are found in γ-LiAlO2 irradiated at 573 K. Complete amorphization with gas bubble formation starts from the surface. Cubic LiAl5O8 precipitates surrounded by cavities form in the damage peak region. Li out-diffusion and vacuum evaporation from the sample surface occur during ion irradiation at elevated temperatures. STEM-EELS provides evidence for preferred pathways of Li diffusion along the grain boundaries. In situ HIM shows volume swelling, He bubble formation in grains and He release via grain boundaries. In situ TEM reveals a rapid formation of cavities due to radiolysis in γ-LiAlO2 under low-energy electron irradiation. The thermal annealing data suggest that the D diffusion coefficients in the irradiated material have a value on the order of 2×10-13 cm2/s at 773 K. There is an exponential decay behavior of D release at room temperature from γ-LiAlO2 pellets irradiated at 188 K; at the low temperature, D atoms appear to be immobilized in the irradiated structure. The data from this study also show significant H and D diffusion and release from γ-LiAlO2 during irradiation at 573 K; D retention increases with increasing D2+ ion fluence and tends to saturate at 4×1017 He++D+/cm2.
9:45 AM - EN17.05.06
Defects, Stoichiometry and Transport Processes in Be12X Intermetallics
Robin Grimes1,Matthew Jackson2,Patrick Burr3,Simon Middleburgh4
Imperial College London1,University College London2,University of New South Wales3,Bangor University4
Show AbstractBe12X (X = Ti, V, Mo and W) intermetallics are a promising family of materials for the first wall and neutron multiplying applications in future nuclear fusion reactors. Due to the difficulties of working with Be they are often poorly characterized. Thus, the crystallography, elastic properties, atomic transport processes, thermodynamical stability and deviation from stoichiometry have been investigated using computer simulation based on density functional theory.
Surprisingly there has been considerable controversy regarding the structure of TiBe12, which is variously reported as hexagonal and tetragonal. Lattice dynamics simulations show the I4/mmm tetragonal phase to be more stable over all temperatures. The formation enthalpies of intrinsic point defects and the elastic constants of all the Be12X intermetallics are then predicted. Defect concentrations were shown to be dominated by antisite disorder and Be vacancies, suggesting these materials can accommodate excess X much more easily than excess Be. In Be12Ti it was found that titanium defects have much higher hopping energy than beryllium across the majority of migration pathways studied. Both beryllium vacancy and interstitial diffusion is weakly anisotropic, however, migration of beryllium divacancies is isotropic with energy equal to that of the isolated vacancy.
EN17.06: Reactor Steels
Session Chairs
Marie-France Barthe
Iuliia Ipatova
Tuesday PM, December 03, 2019
Sheraton, 3rd Floor, Hampton
10:30 AM - EN17.06.01
Influence of Irradiation Conditions on Precipitation Behavior in Fe-Cr and Ni Alloys
Emmanuelle Marquis1,Li-Jen Yu1,Elaina Reese1,Takuya Yamamoto2,G. Robert Odette2,Julie Tucker3,M. Grace Burke4
University of Michigan1,University of California, Santa Barbara2,Oregon State University3,University of Manchester4
Show AbstractThe use of heavy ions, protons, or electrons to understand the effect of irradiation on microstructures and therefore materials properties presents many experimental benefits over the use neutrons. However, extrapolating the laboratory observations to behaviors found under neutron irradiation in reactor conditions is not a straightforward process. In particular, dose rates that are orders of magnitude apart and can play a significant role on the development of microstructures, through mechanisms that include higher point defect production and cascade mixing effects. Here, we will discuss the behaviors under ion, proton, and neutron irradiation of alloys that would normally phase decompose under thermal conditions. Specifically, using atom probe tomography to characterize microstructures at high spatial and chemical resolution, the evolution of microstructures in model ferritic Fe-Cr alloys and austenitic Ni alloys was quantified as a function of dose and dose rates. In Fe-Cr alloys, precipitation of the α’ phase is expected under long thermal annealing times. In the selected commercial Ni alloys (625, 625 Plus, and 690), the γ” phase, an ordered Pt2Mo-type phase, and in some cases the γ’ phase are expected to precipitate. Under irradiation conditions, both alloy systems exhibit different precipitation regimes that are dependent on dose rate. Generally, accelerated precipitation by radiation-enhanced diffusion is observed at lower dose rates, establishment of a steady state regime occurs at intermediate dose rates, and precipitation is entirely suppressed at higher dose rate.
11:00 AM - EN17.06.02
In Situ Micromechanical Testing of Unirradiated and Ion Irradiated Reactor Pressure Vessel Steels
Claudia Gaparrini1,Alan Xu2,Tao Wei2,Ken Short2,Joel Davis2,Tim Palmer2,Dhriti Bhattacharyya2,Nick Riddle3,Lyndon Edwards2,Mark Wenman1
Imperial College London1,Australian Nuclear Science and Technology Organisation2,Rolls-Royce3
Show AbstractRecently there has been a considerable need to extend the lifetime of the world’s pressurised water reactor (PWR) fleet from 40 to 60 or possibly 80 years. The continued safe operation of these reactors is intrinsically linked to the structural integrity of the steel used for the reactor pressure vessel (RPV). Matrix damage and solute clustering induced by neutron irradiation are considered one of the main causes of embrittlement of RPV steel. Intensive research has been ongoing to investigate embrittlement mechanisms of the low alloy ferritic RPV steel, however, testing neutron irradiated materials can take a long time and is very costly. The recent development of micro mechanical testing allows researchers to evaluate mechanical properties of ion irradiated materials which aim to simulate the damaging effects observed on neutron irradiated materials.
This work presents in situ micro tensile testing of a novel manufactured RPV steel: hot isostatic pressed (HIP) SA508 grade 3 steel. Micro tensile specimens with dimensions of 5 x 5 x 15 µm were manufactured using a gallium focused ion beam (FIB) and tensile testing was performed under displacement control at a rate of 20 nm/s in situ in a scanning electron microscope (SEM). Tests were performed under the same conditions on unirradiated and ion irradiated specimens which were irradiated by He2+ ions at 5 MeV at room temperature to the damage of 0.6 dpa using a 2 MV tandem accelerator with energy degrader at ANSTO.
The mechanical properties (in terms of ultimate tensile stress and 0.2% proof stress) of the unirradiated micromechanical tests agreed well with data obtained via conventional mechanical testing. The effects of ion irradiation on the mechanical properties of this RPV steel were measured as an increase of ultimate tensile stress and yield strength via micro tensile testing. The hardening of the irradiated specimens was also measured using nanoindentation performed on the top sample surface. The hardening mechanism can be correlated to the implantation induced features observed on the ion irradiated material that were resolved using a transmission electron microscope. Fine bubbles/voids with an average diameter of 1.2 ± 0.5 nm were observed.
11:15 AM - EN17.06.03
Energetic Drive for Ni, Mn and Si Clustering at Dislocation Loops in bcc Iron
Thomas Whiting1,Daniel King1,Mark Wenman1
Imperial College London1
Show AbstractThere are currently concerns that clusters of Cu, Mn, Ni and Si could limit the operational lifetime of reactor pressure vessels (RPVs) through reductions in fracture toughness, unaccounted for in regulatory models. As there is much debate over the origin of these clusters and whether they are radiation-enhanced or radiation-induced and whether they can be classified as discrete phases, there is an ongoing global research effort to investigate their mechanism of formation.
Using density functional theory (DFT) simulations, we have conducted investigations into the effect of strained environments on the binding energies of solutes, solute pairs, vacancy-solute pairs and dumbbells in bcc Fe to better understand the conditions that favour nucleation and growth of solute clusters as it is expected that these are dependent on both strain fields and the local chemical environment. Atom probe tomography studies have found that clusters are frequently associated with dislocation loops and grain boundaries, suggesting that these could act as nucleation sites for solute clustering. As modelling large dislocation loop sizes is computationally expensive using DFT, we have used interatomic potentials for Fe to model both small (19 atom) and large (65 atom) <111> dislocation loops to identify the range of strain on the lattice. We have then used these results to parameterize DFT simulations by modelling defects in strained lattices that are typical of these loops.
Our findings show that: (1) vacancies will be dragged to these locations as vacancy formation energies are reduced by more than 1 eV when subject to large volumetric strains, (2) solutes will be dragged due to the strong solute-vacancy binding energy for Ni, Mn, Si in both first and second nearest neighbour positions, (3) solutes remain bound to vacancies even in significantly strained environments, and (4) magnetic effects are important to model the energetics of systems, particularly for complex environments such as those including self-interstitial atoms.
11:30 AM - EN17.06.04
Measuring Microstructure Changes in Ion-Irradiated 316LN Stainless Steel
I-Hsuan Lo1,Philip Edmondson2,Karl Whittle1
University of Liverpool1,Oak Ridge National Laboratory2
Show AbstractUK’s nonprescriptive regulatory system has incentivised the nuclear industry to invest in the research and development of reliable structural integrity codes since 1970’s. To support the UK structural integrity services, the UK project on the ‘Multi-scAle INTegrity assessment for Advanced high-temperature Nuclear system (MAINTAiN)’ offers the ability to further strengthen UK’s position in structural integrity through developing a novel computer model-based assessment code. The aim is to develop the knowledge that is necessary to reliably assess the integrity of critical nuclear components that operate in extremes of high-temperature, high mechanical stress, and irradiation. To this end, a predictive three-dimensional multi-scale creep deformation model, informed by the materials' microstructure, will be established. The MAINTAiN project will provide the scientific foundation for the next generation of high temperature structural integrity assessments.
The University of Liverpool has carried out a part of irradiation damage on the dislocation creep behaviour of three materials that are of interest for the next generation of nuclear power plants: Stainless Steel Type 316LN, ODS Steel, and FeCrAl alloys.
The presentation will illustrate the details of the MAINTAiN and the latest results on the irradiation test of 316LN stainless steel. The irradiation damage was induced by self-ion implantation using the Dalton Cumbrian Facility (DCF). In order to investigate changes, this works has included high resolution transmission electron imaging and spectroscopic analysis of damage in order to investigate the changes in irradiation.
11:45 AM - EN17.06.05
Irradiation Assisted Stress Corrosion Cracking in 508-304 Weldment in BWR/NWC Simulated Environment
Brent Heuser1,Zhen Li1
University of Illinois at Urbana Champaign1
Show AbstractReactor pressure vessel steel SA508 (a low-alloying steel) and 304 (Fe-Cr-Ni austenitic stainless steel) are widely employed in both PWRs and BWRs. Weldments typically have high residual stress and composition gradients within the fusion zone (FZ) and heat affected zone (HAZ) that lead to greater stress corrosion cracking (SCC) susceptibility. Irradiation assisted SCC can be controlled by factors including water chemistry, localized deformation, tensile stress, radiation induced segregation and radiation damage. In particular, the redistribution of Cr, the element responsible for oxidation/corrosion resistance via chromia formation, can sensitize these alloys to SCC phenomenon. These materials can be preferentially attacked at active paths, such as GBs in the FZ and HAZ, when exposed to chemically reactive environment and to radiation.
Slow strain rate tests (SSRT) with strain rate up to 10-8/s were applied on proton irradiated EPRI SA508-309 weldment in an autoclave system with simulated the BWR/NWC environment (10.2 Mpa, 288 °C and 2000 ppb oxygen) in order to investigate the radiation assisted SCC. The effect of radiation to induce segregation in the FZ and HAZ of SA508-309 weldment is of typical interest. In addition, SSRTs were employed on as-received SA508-309 weldment to separate the effect of irradiation from environment factors such as LWR water chemistry and applied tensile load. Moreover, SSRTs were conducted on as-received weldment at room temperature and atmosphere pressure. The yield strength and ultimate tensile strength of as-received and proton irradiated SA508-304 weldment under NWC environment is 25% lower compared to that under room temperature and atmosphere pressure. The redistribution of elements such as Cr in the grain boundary of as-received and irradiated weldment is investigated with STEM-EDS. The initiation of cracking was studied by SEM and TEM in the as-received and irradiated weldment. This talk will focus on identification of the primary factors of SCC in Fe-based weldments.
EN17.07: Molten Salt Corrosion and Chemistry in Advanced Nuclear Reactors
Session Chairs
Chaitanya Deo
Simon Middleburgh
Tuesday PM, December 03, 2019
Sheraton, 3rd Floor, Hampton
1:30 PM - EN17.07.01
Studying Corrosion in Materials for Nuclear Applications Using Synchrotron-Based X-Ray Methods
Simerjeet Gill1
Brookhaven National Laboratory1
Show AbstractDevelopment of next generation materials for nuclear energy structural applications requires understanding structural changes and failure mechanisms under extreme environments. Reactions at interfaces under corrosive environments play a crucial role in material failures in current and advanced reactors.
Our research focuses on utilizing synchrotron characterization techniques such as spectroscopy, multi-modal imaging and diffraction for characterizing material interfaces and their properties when subjected to extreme environments of high temperature and radiation fluxes present in nuclear reactors. Synchrotron based characterization studies performed for structural materials and molten salt systems, elucidating corrosion at interfaces will be presented.
X-ray absorption spectroscopy has been utilized to investigate local coordination environment, bonding dynamics and electronic properties of metal species, and their changes at elevated temperatures in molten salt systems. Using a custom-designed in situ cell, we used Ni and Co dopants in single component ZnCl2 and ZnCl2-KCl eutectic and studied their properties using Ni, Co and Zn K-edge spectra. The local structural changes and speciation of metallic species in molten salt systems will be discussed.
3D multi-modal imaging techniques will be discussed to provide quantitative characterization of Cr segregation at grain boundary inside the sensitized steel sample using X-ray Fluorescence. Tomographic image reconstruction of stainless steel deteriorated by IG corrosion is combined with 3D imaging of phase contrast using Differential Phase Contrast imaging. This combined structural and chemical analysis of the Cr segregation at grain boundaries allows us to understand IG corrosion in sensitized stainless steel at nanoscale.
To track degradation of materials in real time, we built and commissioned an in situ sample environment for studying interfaces under high-temperature and pressure conditions using x-ray diffraction to investigate corrosion mechanisms in cladding materials. Capabilities of the in situ sample environment will be discussed.
2:00 PM - EN17.07.02
First Principles Investigation of Cr Segregation Behaviors in Ni-Cr Alloy in Molten Salt Systems
Jacob Startt1,Stephen Raiman2,Chaitanya Deo1
Georgia Institute of Technology1,Oak Ridge National Laboratory2
Show AbstractThe development of new structural materials that can withstand the degradative and corrosive environments exhibited by molten salt systems currently stands among the most important remaining scientific targets that must be reached before viable molten salt systems can be realized at an industrial scale. In corrosion focused studies, the prevailing trend observed in the most commonly investigated alloys (such as Hastelloy N or 316 stainless steel) is the depletion and dissolution of Cr atoms near the alloy surface and along grain boundaries. The atomic level mechanisms and forces that drive the migration of Cr and its eventual removal from the surface are not well understood, and due to the harsh environments intrinsic to molten salt systems experimental efforts may struggle to observe these fundamental processes. Carefully devised computational modelling offers another avenue to study the nature of these interactions and may even help to suggest potential mechanisms and to narrow the field of possible reaction pathways for future experimental studies to investigate.
In this study, the segregation behavior of Cr near an (100) fcc Ni surface is investigated with first-principles density functional theory (DFT), first for a clean surface and then for surfaces with various salt species adsorbed above the segregating Cr atom. Surface segregation energies describe the energetics associated with the location of a solute atom near a surface (i.e. the relative energy of an atom when its located in the bulk versus in a specific surface layer).
Calculations show that under a clean surface (i.e. no adsorbed salt atoms) it is energetically favorable for the Cr atom to inhabit the 3rd or 2nd surface layer, but largely unfavorable for it to sit in the topmost 1st layer. When a Cl adatom is placed above the Cr atom, the segregation energy for the top-most layer decreases, suggesting Cl surface adsorption lowers the barrier for Cr to move to the top layer. When F or O are adsorbed to the surface, the segregation energy not only decreases but becomes significantly negative, suggesting that Cr segregation to the top layer not only becomes more favorable with these species on the surface but also actually preferred relative to a remaining in a bulk position. Further investigation into the effects of H-Cl, H-F, and H-O will also help to understand the role of H and H2O salt impurities in the corrosion process. The additional presence of an H atom may also allow the highly electronegative surface adsorbates to increase their local electron occupation level, which would more closely resemble the charged ionic nature exhibited by the salt in a molten state. This talk will report new insights into material-salt interfaces through atomistic modeling.
2:15 PM - EN17.07.03
Corrosion Mechanism of Molten Salts on Haynes 230 Alloy Studied Using In Situ Neutron Reflectometry
Joohyun Seo1,Mathieu Doucet1,Gabriel Veith1,Sheng Dai1,James Browning1
Oak Ridge National Laboratory1
Show AbstractMolten salt eutectics are used as heat transfer fluid in concentration solar power systems. One of the technological challenges with such systems is the control of corrosion of the materials used for transport of the molten salts. For this reason, understanding how the interface between salts and alloys change as a function of temperature is important. In this study, we focused on the interfacial structure at the interface of KCl-MgCl2 molten salt eutectic and Haynes 230 alloy using neutron reflectometry. This binary molten salt eutectic was deposited on Haynes 230 alloy to characterize the multilayer as a function of temperature. Our measurements were carried out in situ under ultra-high vacuum conditions. First, a thin film of a single Haynes 230 alloy layer was studied to characterize the alloy as a function of temperature before exposing it to molten salts. The structure of the Haynes 230 alloy film was measured between room temperature to 451 °C. Subsequent measurements were made with a thin film of salt sandwiched between two layers of Haynes 230 alloy at temperatures below the salt’s melting point with sample temperatures ranging from ambient to 383 °C. In both cases, a change in the Haynes 230 alloy structure is observed at around 160 °C. We will describe the environment chambers used for these measurements as well as our results of the behavior of the interfacial structure in these thin film systems.
2:30 PM - EN17.07.04
Tritium Mobility and Local Chemistry in Fluoride Molten Salts
Stephen Lam1,Ronald Ballinger1,Ju Li1,Charles Forsberg1
Massachusetts Institute of Technology1
Show AbstractMolten fluorides have been proposed for use as a coolant in various advanced nuclear concepts including molten-salt reactors (MSRs), fluoride high temperature salt-cooled reactors (FHRs) and fusion devices. In these systems, significant quantities of tritium will be produced which must be captured and controlled to prevent a radioactive release. In order to efficiently remove tritium from the system, the transport and chemical properties must be well characterized and understood. Yet, difficulty in the handling of fluoride salts and tritium has resulted in relatively few experiments being conducted and large discrepancies across different studies. In this work, the transport and local chemistry of different tritium species (tritium fluoride and diatomic tritium) has been studied in the prototypical fluoride salts FLiBe (66.6-33.3 mol% LiF-BeF2) and FLiNaK (46.5-11.5-42 mol% LiF-NaF-KF) using ab-initio molecular dynamics (AIMD). The methods were validated using experimentally well-characterized fluorides. Salts examined in this study include binary fluorides LiF, NaF, and NaF, and ternary fluorides LiF-BeF2 and LiF-KF at various compositions. Radial distribution functions and coordination numbers from neutron diffraction, and diffusivities and activation energies from nuclear-magnetic resonance are accurately reproduced. Chemical species and structures were found by using graph-theoretic representations of the system at each time step and removing noise with probabilistic filtering methods. Over the temperature range from 873 to 1373K, the T2 and TF diffusivities in FLiBe were found to be D[m2/s]=4.14E-7*exp(-33.0[kJ/mol]/RT) and D[m2/s]=1.724E-7*exp(-32.0[kJ/mol]/RT) respectively. At the same temperatures in FLiNaK, T2 and TF diffusivities were D[m2/s]=1.38E-6*exp(-39.0[kJ/mol]/RT) and D[m2/s]=4.94E-7*exp(-41.0[kJ/mol]/RT) respectively. In both systems, TF exhibited 3-4 times lower diffusivity than the reduced form of tritium T2. In both cases, decrease in diffusivity due to loss in mobility is attributed to TF reactivity and complexation in the salt. In FLiBe, TF was found to be bound to various BeF2 molecules, while in FLiNaK, TF2- complex was formed. In contrast T2 molecules in both simulations were found to be stable in pure salt with limited chemical interaction in the solvent. The tritium diffusivities fall within range of experimental values and discrepancies between different experimental studies in literature were attributed to a combination of the following factors: differences in measured species (TF and T2), presence of impurities (corrosion products, moisture) and differences in isotopes used (H, D, T). Large differences in tritium transport will have significant impact on assessing tritium control strategies, which typically rely on mass transport (adsorbents, permeation windows) and chemical condition (redox controls). This study provides a deeper understanding of chemical interactions and accurate chemical-transport data that can be used to predict macroscopic behavior and design engineering and control systems.
2:45 PM - EN17.07.05
High Temperature Cell for In Situ Reflectometry of Molten Salts
Mathieu Doucet1,James Browning1,Joohyun Seo1,Gabriel Veith1
Oak Ridge National Laboratory1
Show AbstractNext generation concentrated solar power systems and molten salts reactors have created new interest in molten salt chemistry. The corrosion processes of the alloy used to contain the molten salts as a function of temperature remains to be fully understood. In particular, we are interested in the corrosion of Haynes 230 (approximately Ni>57 Cr22 W14) in contact with a KCl/MgCl2 eutectic mixture for application in solar power systems. Our team has focused on studying this corrosion process in situ using neutron reflectometry. We will present the design of the high temperature cell that will enable us to perform those in situ measurements from room temperature to 850 C.
EN17.08: Zirconium Cladding
Session Chairs
Simerjeet Gill
Joohyun Seo
Tuesday PM, December 03, 2019
Sheraton, 3rd Floor, Hampton
3:30 PM - EN17.08.02
Free-Energy Functional of the Zirconium-Hydride System from First Principle Calculations
Michele Fullarton1,Simon Phillpot1,Yongfeng Zhang2,Larry Aagesen2
University of Florida1,Idaho National Laboratory2
Show AbstractZirconium based alloys for nuclear fuel cladding are known to precipitate hydride phases during operation. These precipitates can cause several issues including embrittlement leading to cracking and failure of the cladding. In order to predict the materials properties of the cladding the hydride phases must be accurately simulated. In this work the thermodynamic stability of phases in the zirconium-hydride system was evaluated using first principles-based methods. Cluster expansion techniques were employed to assess the stability of each phase over a range of concentrations at zero temperature. Phonon contributions for each ground state were calculated and combined with results of finite-temperature Monte Carlo simulations to create Gibbs free energies of the phases as a function of temperature and concentration. From these results a new phase diagram was generated, which will clarify discrepancies between different experimental phase diagrams. Moreover, the free energy functionals of all relevant phases were parameterized, allowing for more accuracy calculating microstructure and materials properties of cladding, for instance, in phase field simulations with INL’s MARMOT code.
3:45 PM - EN17.08.03
Oxygen Stoichiometry Deviation in Amorphous ZrO2 and Yttria-Stabilized Zirconia
Simon Middleburgh1,Michael Rushton1,Iuliia Ipatova1,Lee Evitts1,William Lee1,2
Bangor University1,Imperial College London2
Show AbstractAmorphous zirconia (a-ZrO2) has been modelled using a combination of reverse Monte-Carlo, molecular dynamics and density functional theory together. This combination has enabled the complex chemistry of the amorphous system to be efficiently investigated. Interestingly, the a-ZrO2 system was observed to accommodate excess oxygen readily through the formation of neutral peroxide (O22−) defects – a result that has implications not only in the a-ZrO2 system, but also in other systems employing network formers, intermediates and modifiers. Similar conclusions are made for the yttria-stabilized zirconia system and has been corroborated with Raman spectroscopy data. These results have potential implications for materials behaviour in reactor systems: notibly a mechanism for oxygen to be accommodated and transported through thin, amorphous zirconia films such as those expected at highly damaged grain boundaries - providing a mechanism for zirconium cladding corrosion.
4:00 PM - EN17.08.04
Deformation and fracture of Zirconium Hydrides During Plastic Straining of Zr-4
Luca Reali1,Said El Chamaa1,Daniel S Balint1,Mark Wenman1,Adrian Sutton1
Imperial College London1
Show AbstractCrack initiation in zirconium alloys is an important issue affecting the safety of nuclear reactors. Zirconium hydrides that precipitate in service are potential crack nucleation sites. In this work, the deformation and cracking of zirconium hydrides was studied during the deformation at room temperature of Zircaloy-4 samples up to fracture. Two hydrogen concentrations of 100 and 200 ppm were considered.
Hydrides precipitate both inside and at the boundary of the grains. It is expected that thicker hydrides and hydrides at the grain boundary are more prone to fracture. The main objective of this study was to better understand which of the two is the most important factor, and to identify at which point in the deformation of the alloy the first hydrides break. Microvoids thus nucleated may coalesce and be an important factor in understanding the fracture of these alloys. Scanning microscopy (SEM) images of a number of hydrides, both intergranular and intragranular, were taken at discrete increments of deformation, and the fraction of those that show evidence of fracture was estimated.
The results show that intergranular hydrides possess a lower deformability, which can be rationalised by analysing the conditions for slip transmission into and out of the hydride.
4:15 PM - EN17.08.05
Hydrogen Pickup and Oxidation Kinetics of Zirconium Alloy—Dopant Effects from the Chemo-Mechanical Perspective
Jing Yang1,Mostafa Youssef1,Bilge Yildiz1
Massachusetts Institute of Technology1
Show AbstractEffect of alloying elements on zirconium alloy degradation properties has been widely studied because of its application in nuclear water reactors. It is observed in experiments that alloying element in zirconium alloy could result in different morphology of the passive film as well as different oxide film growth kinetics. The goal of this work is to study oxygen diffusion and hydrogen pickup processes in the ZrO2 passive film system under the effect of alloying elements, with varying temperature and chemical environment. In particular, we focus on the chemo-mechanical effect of Fe, Cr, Sn and Nb dopants in ZrO2 that exist in the oxide matrix as point defects from first-principles calculations. We show that the change in defect chemistry imposed by doping is strongly correlated with the valence states of the dopants. Fe and Cr, which are dominantly in +3 valence states, tend to decrease the concentration of oxygen interstitials and increase the concentration of hydrogen interstitials in ZrO2, thus suppressing oxide growth while increasing hydrogen pickup. Nb, on the other hand, is dominantly +5 and has the opposite effect. Changes in defect chemistry are accompanied by change in phase stability. We have discovered that Fe and Cr can strongly stabilize tetragonal-phase ZrO2 while Nb has a strong destabilizing effect. This effect provides insight for the different oxide film morphology of zirconium alloys with varying alloying element.
4:30 PM - EN17.08.06
Influence of Boron Isotope Ratio on the Thermal Conductivity of Uranium Diboride (UB2) and Zirconium Diboride (ZrB2)
Simon Middleburgh1,Lee Evitts1,Erofili Kardoulaki2,Michael Rushton1,Iuliia Ipatova1,William Lee1,3
Bangor University1,Los Alamos National Laboratory2,Imperial College London3
Show AbstractUranium diboride is being considered as an advanced fuel, while zirconium diboride remains a leading candidate for use as a burnable absorber, often coated on uranium dioxide fuel pellets. To ensure safe and efficient operation models need to be developed to enable the material to be modelled within a fuel performance code. Thermal properties of these materials are therefore investigated using density functional theory methods. It was found that the boron isotope ratio significantly varied the thermal conductivity for both materials. When compared to experimental data, the thermal conductivity for ZrB2 is well prediced, however, a significant discrepancy is observed for UB2. Potential reasons for the discrepancy, including the Hubbard factor and phonon relaxation times, are explored and discussed.
EN17.09: Poster Session
Session Chairs
Chaitanya Deo
Maik Lang
Simon Middleburgh
Maria Okuniewski
Wednesday AM, December 04, 2019
Hynes, Level 1, Hall B
8:00 PM - EN17.09.03
Detection of Low Dose Gamma-Ray Using Color Change of the Organic Conjugated Molecules
Seung-Hwan Oh1,Jin-Moon Yun1,Hyun Bin Kim1
Korea Atomic Energy Research Institute1
Show AbstractRadiation technology is widely used medical radiation therapy, radiation dosimeter, food preservation and other industrial applications. These include scientific research such as high energy physics, radiochemistry, change of material properties and so on. Radiation detection is one of important research area to safely use ionizing radiation. Several devices are investigated for radiation detection such as Silicon diode, metal-oxide semiconductor field effect transistors, thermo-luminescent dosimeters, optically stimulated luminescence dosimeters. These devices use semiconducting properties of organic and inorganic materials with electrical or optical connection for readout.
In this study, we synthesized organic conjugated molecule for application to detection of gamma-ray without any readout devices. This molecule is consisted of two main parts. One is low-band gap conjugated core part, which is dithiophene moiety, for wide absorption in visible range. And the other is dicyanomethylene parts which are possible to induce band-gap change of dithiophene core part by gamma-ray. After irradiation gamma-ray with 1 Gy on film of the molecule, dark blue color is changed to yellow color due to change of band-gap from 1.74 eV to 2.16 eV. We investigate optical and electrical properties of the molecule by gamma-ray and we also discuss detail mechanism of color change and effect on the molecule by gamma-ray.
8:00 PM - EN17.09.04
Anode Material Development for Scale-Up Oxide Reducer in Pyroprocessing
Sung-Wook Kim1,Min Ku Jeon1,Sang-Kwon Lee1,Eun-Young Choi1
Korea Atomic Energy Research Institute1
Show AbstractPyroprocessing has been investigated as an option to recycle and to manage spent nuclear fuels (SNFs) from pressurized water reactors. The oxide SNFs needs to be transformed to metallic states for electrochemical recovery of transuranic elements. In this respect, oxide reduction (OR) technique has been developed for the metal conversion. LiCl molten salt-based electrochemical reaction is the common OR technique. Conventional OR process is designed based on Li2O recycle reaction. Li2O is electrolyzed to form Li metal at cathode and O2 gas at anode (2Li2O = 4Li + O2(g)). The Li metal subsequently reduces the oxide SNFs (ea.g., UO2 + 4Li = U + 2Li2O) to release new Li2O. Hence, Li2O concentration in the LiCl salt is expected to be constant during the OR operation. Pt has been widely used as the O2-evolving anode material because of its superior mechanical and chemical stabilities in the high-temperature oxidation environment.
Scale-up is an important issue on feasibility examination of pyroprocessing. The Pt anode, however, is not preferred in the large-scale system because it is gradually damaged upon the operation resulting in high processing cost. The Pt anode is known to have three anodic reactions depending on anode potential: Li2PtO3 formation (2Li+ + Pt + 3O2- = Li2PtO3 + 4e- at low potential), O2 evolution (2O2- = O2(g) + 4e- at medium potential), and anodic dissolution (Pt = Pt2+ + 2e- at high potential). Only the O2 evolution is required and thus the operation voltage of the Pt anode system should be carefully controlled (normally, ~3 V). Nevertheless, it was revealed that the side reactions cannot be avoided to induce the Pt anode deterioration, even in small-scale experiments (~15 g-UO2). This would be due to non-uniform potential distribution across the Pt anode (by iR drop and other reasons) and local inhomogeneity of the Li2O concentration in the salt. This should become more severe with the scale-up and it is considered that the precise control of the reaction pathway of the Pt anode would be very difficult in the large-scale equipment.
KAERI (Korea Atomic Energy Research Institute) has examined several materials to replace the Pt anode. One approach is to develop the O2-evolving anode with better electrochemical stabilities using conductive ceramics and other noble metals. C anode was also suggested. The C anode system operates at high voltage range (>4 V) compared to the Pt anode system to decompose the LiCl salt to generate Cl2 gas (2Cl- = Cl2(g) + 2e-) at the C anode. The advantages of the C anode are non-concern on the anodic reaction control and high operation current. In this presentation, the recent studies on the anode material development done at KAERI will be introduced.
8:00 PM - EN17.09.05
Interaction Behavior between U-Zr Alloy Containing Lanthanides and Reusable Crucible Materials
Seoungwoo Kuk1,KyungChai Jeong1,SeokJin Oh1,KiHwan Kim1,Jeong-Yong Park1
KAERI1
Show AbstractA U- 10 wt.% Zr alloy containing lanthanide elements (Nd, Ce, Pr, La) was introduced as a surrogate for transuranium alloys in metallic nuclear fuel. U-Zr alloys containing the lanthanide elements interact significantly with graphite crucibles. Conventional coating material, Y2O3 reduced the interaction, but partially interacted and detached from crucible after casting due to the difference in the thermal expansion coefficients of the crucible and the coating layer. The crucible was contaminated and could not thereafter be reused. Thus, it was considered a form of radioactive waste. Moreover, interacted materials could not be removed from the casting part and amounts of the reaction reduced casting yields. Materials such as LaYO3, SiC, TiC, and TiN along with density-improved Y2O3 were considered as candidates for use in the creation of a reusable crucible, which could be used to improve casting yields and to reduce radioactive wastes. The interactive behaviors of the candidate materials were considered by means of a sessile drop test. The annealing temperature was set to 1,450 °C for investigating interactions of the casting method. Microstructures of the candidate materials were observed using a scanning electron microscope (SEM), and their elemental distributions were characterized using energy-dispersive X-ray spectroscopy (EDS). Phase formations of the candidate materials were identified using an X-ray diffractometer (XRD). The material interaction was reduced by the enhanced Y2O3 coating and LaYO3. A shortlist of candidate materials for use in the creation of a reusable crucible was drawn up based on the results of this research.
8:00 PM - EN17.09.06
Reaction Characterization between Graphite Crucible and U-Zr Melt
Seoungwoo Kuk2,Seong-Jun Ha1,2,Jeong-Yong Park2,KyungChai Jeong2,Sang Gyu Park2,Young-Mo Ko2,Yoon-Myeng Woo2,Young-Kook Lee1
Yonsei University1,Korea Atomic Energy Research Institute2
Show AbstractU-Zr metallic fuels have been developed as a nuclear fuel for Sodium-cooled fast reactor because metallic fuels are related to excellent reactor safety and fuel cycle economy. In the fabrication process of U-Zr metallic fuels, the graphite crucible is commonly used for melting and casting and yttrium oxide is coated on the inner surface of the graphite crucible to prevent the reaction with melted U-Zr alloy. It is crucial to prevent the reaction between the melt and crucible for controlling the loss in the fabrication. In this study, we investigated the reactivity between uranium, zirconium and graphite crucible. The graphite crucible was used to clarify how the graphite react with melt uranium and zirconium. U-10wt.%Zr alloy was melted using injection casting apparatus which is an induction furnace. Raw materials consisting of zirconium sponge and uranium metal were charged in the graphite crucible in the order named. The materials were superheated at 1,600°C for alloying and it was cooled in the furnace after alloying. Microstructures in the surface region of the melt-residue of U-Zr alloy were characterized using Scanning electron microscopy (SEM) equipped with Energy dispersive X-ray spectroscopy (EDS). Although the graphite crucible was used for melting uranium and zirconium, the U-10wt.%Zr melt residue was fabricated without alloying between uranium and the graphite crucible. The melt residue was easily separated from the graphite crucible. However, the reaction layer was formed on the surface of U-10wt.%Zr melt residue. The reaction layer mostly consisted of carbon and zirconium. The thickness of reaction layer was about 5 μm at side region and 10 μm at bottom region. The bottom region had the thickest reaction layer among whole surface because first of all, zirconium was reacted with graphite crucible at the bottom region when the charged raw materials were heated. The thickness of the reaction layer was assumed as the depth penetrated into the U-Zr matrix from the graphite crucible depending on the volume of the zirconium, melting temperature and time. The surface of melt residue was investigated using X-ray diffractometer (XRD). As a result of XRD analysis, ZrC and U3O7 were detected on the reaction region. It is likely that C was contaminated from graphite crucible. The quantitative chemical composition of graphite crucible was investigated using Inductively coupled plasma-atomic emission spectroscopy (ICP-AES). AS a result of ICP-AES analysis, uranium and zirconium were contaminated at contact region. On average, uranium and zirconium contents of graphite crucible were 281 and 31 ppm but it was a considerable difference according to the position. In this study, U-10wt.%Zr alloy was fabricated in the optimizing manufacturing process to examine the reactivity between uranium, zirconium and graphite crucible. The reaction layer consisted of zirconium and carbon is formed on the whole surface of the melt residue because zirconium is commonly reacted with carbon at the contact region with graphite crucible. The graphite crucible is also contaminated with uranium and zirconium when melting.
8:00 PM - EN17.09.07
Combined Ab Initio and Empirical Model for Irradiated Metal Alloys with Focus on Uranium Alloy Fuel Thermal Conductivity
Shuxiang Zhou1,2,Ryan Jacobs1,Dane Morgan1
University of Wisconsin-Madison1,Idaho National Laboratory2
Show AbstractHigh-quality thermal conductivity data is critical to the rational design of materials for applications ranging from thermoelectrics to nuclear reactors. However, such data is experimentally and computationally challenging to obtain due to any or all of cost, technical hurdles, and time requirements. The thermal conductivity of these materials can be further influenced by their working conditions, e.g. irradiation, due to the microstructure changes. Therefore, practical models and quantitative understanding of the thermal conductivity, including changes due to microstructure, is essential to support the development of new materials.
In this work, we developed a computational model of thermal conductivity of irradiated metal alloys based on density functional theory (DFT) calculations, semi-classical scattering theory, and experimental data. This approach has been applied to model the electrical and thermal conductivity of alpha uranium previously, and can be extended to model the condictivities of U-alloys like U-Zr and U-Mo. The present model incorporated the contributions of point defects (vacancies and transmutation products), grain boundaries, and noble gas bubbles (inter- and intra-granular) resulting from irradiation. The effects of point defects and grain boundaries are modeled by both DFT calculations and semi-classical scattering theory, and the effect of noble gas bubbles are modeled using empirical models. The utility of our model is demonstrated on U-Mo alloys, which are promising nuclear fuel materials for use in next generation nuclear reactors, and displays semi-quantitative agreement with experimental data with a root-mean-square error of about 0.8 W/m-K (≈7% error versus experiment) over the typical operating temperature and burn-up range for U-Mo alloys. This work provides insight into the semi-quantitative relationships of the thermal conductivity changes resulting from the different irradiation-induced changes in atomic structure and microstructure over a wide range of temperatures, and shows that scattering of electrons from point defects produce the majority of the reduction of the thermal conductivity of U-Mo alloys under irradiation. As an example, at 50% U235 burn-up, the thermal conductivity decreases about 20% due to the effect of point defects, while grain boundary and gas bubble formation contribute only about another 10% decrease. Furthermore, our model can be straightforwardly applied to different irradiation conditions or microstructures by changing the values of of a few corresponding parameters.
The present model both serves as a powerful tool for quantitative semi-empirical modeling of thermal conductivity in irradiated metals, and for materials design insight for the control of thermal conductivity in metal alloys in applications experiencing irradiation, such as the further development of advanced metallic U-alloy nuclear fuels.
8:00 PM - EN17.09.08
Effects of Initial Microstructure on the Hot Extrusion Properties of Annular Fuel after Extrusion
Sang Gyu Park1,Seoungwoo Kuk1,Jeong-Yong Park1
KAERI1
Show AbstractThe innovative fuel development is the development of the advanced sodium-cooled fast reactor metallic fuel concepts. The fabrication experiment seeks to investigate advanced fuel designs with the following features: decreased fuel smeared density, venting of the fission gas to the sodium coolant, reduce the FCCI (Fuel Cladding Chemical Interaction), and an advanced fabrication method that includes consideration of annular fuel and extrusion method. The one of most attractive advantage of extrusion method is save the process waste by omitting the sodium process. From the previous study, annular fuel shows the possibility of the reduction of swelling effect and then prevention of the FCMI (Fuel Cladding Mechanical Interaction). However, the fabrication technology of the annular fuel has not been developed yet. Hence, KAERI is developing the extrusion type metal fuel manufacturing technology as a part of the development of the original technology for the production of innovative metal fuel. In this study, the prototype of annular fuel has been fabricated by using Cu billet. The design of billet and annular fuel has been determined, and then and design and material for the mold has been determined by using Deform 3D program. After the mold fabrication, the prototype annular fuel has been fabricated and its texture were examined by us EBSD (Electron Back Scatter Diffraction). There are two types of billets were used for the extrusion in this study; one is drawn billet which has very fine grains(less than 100um), and the other is casted billet that has very coarse grains (<500um). When we compared the grain size after extrusion, the microstructure of the drawn billet showed a grain size of at most 100um or more, which is relatively coarser than that of the extruded annular fuel. Therefore, it is considered that the non-anisotropic crystal tendency analyzed after extrusion production is attributed to the fact that the crystal grains used in the processing billet are already small in size. On the other hand, the texture of annular fuel from the casted billet shows very coarse and anisotropic tendency after extrusion. Therefore, the control of initial texture of billet seems the main factor for the annular fuel microstructure. The fabrication methods for the billet microstructure control will be discussed.
8:00 PM - EN17.09.10
Determination of Dose Effects on Defect Accumulation under Irradiation in Nanoporous Gold and Niobium via Atomistic Simulations
Chaitanya Deo1,Daniel Vizoso1,2,Remi Dingreville2
Georgia Institute of Technology1,Sandia National Laboratories2
Show AbstractThis study explores the radiation resistance of metallic nanofoams and nanowires via atomistic simulations. Nanostructured metals present a high density of interfaces and surfaces as sinks for radiation produced point defects and thus may offer a means to developing radiation tolerant materials. Current work in the modeling of radiation damage in nanofoams and nanowires typically considers defect accumulation due to a small number of primary knock on atoms. For this work, dose effects have been investigated by varying the number and energies of primary knock on atoms, including the consideration of damage accumulation due to damage events with a spectrum of energies. The study investigates the impact of structural properties such as ligament length and diameter on defect accumulation mechanisms and defect properties, such as formation and binding energies as well as the relationship between energies of the primary knock on atoms and changes in mechanical properties. This analysis is performed in two different nanoporous material systems: gold and niobium. The radiation response of nanoporous gold has been studied both experimentally and computationally, although previous computational works have been limited in the doses considered. Nanoporous niobium is a material system that has been previously synthesized but not had its radiation response characterized either computationally or experimentally. The comparison of the radiation responses of nanoporous gold with that of nanoporous niobium also allows for an investigation into the differences in behavior between bcc and fcc nanoporous media. Results suggest that the presence of interfaces and surfaces in nanofoams affects the formation and migration of defect clusters leading to a more radiation resistant metallic structure.
Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy National Nuclear Security Administration under contract DENA0003525.
8:00 PM - EN17.09.11
Reusable Multifunctional 3D Graphene Electrode for Highly Efficient In Situ Extraction of Uranium from Mining and Contaminated Solution
Ahmed Elwakeil1,Chao Wang1,Ju Li1
Massachusetts Institute of Technology1
Show AbstractThe potential strategic application of uranium in nuclear power industry, as well as the environmental and human health problems caused by uranium, promotes the development of reliable methods for in situ extraction of U (VI) from different solutions. Herein, a unique multifunctional three dimensional electrodes have been fabricated by simple hydrothermal method and used as outstanding electrosorbent for uranium extraction from different aqueous solution. The electrochemical sorption method solves the conventional drawbacks of other methods by using fixed potential to guide the movement of uranium ions in the solution and increase the collision rate onto the electrode surface, resulting in faster kinetics and higher adsorption capacity, easier scaling up, and high selectivity due to specific grafted functional groups. The fabricated electrode was characterized by RAMAN, FTIR, XRD, STEM, EDS, XPS before and after uranium loading. The adsorption capacity for U(VI) ions with our multifunctional 3D electrode is estimated to be 4800 mg/g at pH 5.2 without reaching saturation exhibiting the highest reported sorption capacity compared with other electrochemical methods. The fabricated electrode can be efficiently regenerated and recycled for at least 7 times in uranium solution with concentration 40th times higher than initial concentration without any observable change in efficiency. This work represents the first application of multifunctional 3D electrode for uranium extraction from mining solution and contaminated ground water.
8:00 PM - EN17.09.13
Glass Degradation and pH Changes in Narrow Openings of Cracks
Rama Krishna Chinnam1,2,William Lee2,3
University of Limerick1,Imperial College London2,Bangor University3
Show AbstractIn an earlier degradation study on partially immersed international simple glass (ISG), it was found that water near the surface of the ISG or localised variation in pH resulted in discrepancy of glass dissolution and variation in precipitates size on the glass surface. Another finding in the study was the formation of pits on the glass surface under the gel layer [1].
In continuation to the above study, new experiments were designed to understand and measure the pH changes in the localised regions of the glass. For this purpose
1. A study was performed to understand the influence of confined narrow opening on the glass degradation. This experiment helps to understand the environment between gel layer and glass, and to simulate the degradation conditions inside a glass crack.
2. A study was performed to understand the change in pH of the water inside a glass crack with access to fresh water and glass cracks without access to fresh water.
In 1st study, two ISG slides were stacked, an excessive gel layer was found at the outer rim of the interface and the gel layer was found absent at the centre of the interface with a clear boundary between the two regions. This effect was studied under the electron microscope and the TOF-SIMS to confirm the chemical composition difference between the two regions. Thus indicating that degradation in narrow openings of glass is quite different.
In 2nd study, water inside cracks with access to fresh water was found to exhibit pH 10 and pH 8. Water in the cracks with no access to fresh water was found to exhibit pH 11, pH 12 and pH 13. The change in the pH of water in cracks is related to the depth of the crack and position in the water. Thus confirming that cracks exhibit different pH conditions than surrounding water, fresh water availability and their position within water.
[1] RK Chinnam, PCM Fossati and WE Lee, “Degradation of Partially Immersed Glass: A New Perspective,” J. Nuclear Mats. 503 56-65 (2018).
8:00 PM - EN17.09.14
Investigations in the UB2-UO2 Advanced Fuel System
Fabio Martini1,Iuliia Ipatova1,Lee Evitts1,William Lee1,Michael Rushton1,Antoine Claisse2,Simon Middleburgh1
Bangor University1,Westinghouse Electric Sweden AB2
Show AbstractUranium borides are attractive candidates for advanced nuclear fuels. In particular, uranium diboride-based fuels are expected to outperform – in terms of thermal conductivity and spatial uranium density – their currently-used counterparts based on uranium dioxide. Nevertheless, uranium diboride is expected to be less tolerant than uranium dioxide upon accidental contact with water and other oxidizers at high temperatures. The dispersion of uranium diboride in a uranium dioxide matrix is seen as a possible solution that combines the advantages of both compounds, affording a composite material that has a greater thermal conductivity and spatial uranium density than pure uranium dioxide while maintaining its resistance to oxidation.
The interface chemistry of UO2-UB2 systems has been investigated via temperature dependent atomic scale simulations.
8:00 PM - EN17.09.15
Multifunctional Nanoceramic Barrier for DEMO Breeding Blanket Concepts
Matteo Vanazzi1,2,Boris Paladino1,2,Daniele Iadicicco1,Patricia Munoz3,Teresa Hernandez3,Serena Bassini4,Marco Utili4,Fabio Di Fonzo1
Istituto Italiano di Tecnologia1,Politecnico di Milano2,CIEMAT3,ENEA4
Show AbstractThe realization of the breeding blanket system represents one of the crucial points in the design of future generation fusion reactors. At the time being, issues related to materials compatibility persist. According to the most relevant design, fusion reactors will take the Tritium-Deuterium fusion as the reference reaction for power generation. Thus, the availability of Tritium to fuel the reactor core assumes relevant importance. The breeding process represents one of the main focus points of technological R&D activities and the inhibition of Tritium permeation is mandatory to achieve Tritium balance in the reactor chain: once Tritium is produced, an adequate permeation barrier is required to confine it. At the moment, two are the eligible breeding materials under investigation: a flowing Pb-16Li eutectic alloy and a pebble bed composed by sintered Lithium ceramics and Beryllium. It was demonstrated that both these two materials lead to severe corrosive attack of the steel constituting the breeding blanket moduli (RAFM steel), hence threatening the mechanical integrity of the whole structure. It is thus clear that materials are proving to be one of the major bottlenecks for this technology. In order to undertake these issues, we report on multifunctional nanoceramic oxide coatings grown by Pulsed Laser Deposition (PLD) technique. Oxides were chosen because of their chemical inertia and thermodynamic stability. Among these, Aluminium Oxide and Yttrium Oxide were identified as promising candidates. Deposited films appear dense, compact and well adherent on the steel substrates, showing optimal metal-like mechanical properties combined with high hardness. Morphological and crystalline features were evaluated in the case of as-deposited and annealed samples. A deep characterization was performed by Scanning Electron Microscopy (SEM), Energy Dispersive X-ray (EDX), Fourier Transform Infrared spectroscopy (FTIR) and X-ray Diffraction (XRD). Also, coatings of these materials have been tested as Tritium permeation barriers with Hydrogen at different temperatures (from 350°C to 650°C). They showed a permeation reduction factor (PRF) up to 105 at 650°C. These results have been confirmed also in the case of Deuterium permeation, both under 1.8 MeV electron irradiation and for samples that were previously irradiated with ions. In addition, to evaluate the chemical compatibility of the films with the selected breeding materials, samples have been exposed to corrosion tests (both Pb-16Li and Lithium ceramic pebbles) up to 10000 hours. No corrosive attacks on the steel substrate were detected. Finally, Atomic Layer Deposition technique has been studied and recently employed as a complementary deposition technique to the PLD. The deposition process is still at its early stages of development, but preliminary results show that the coatings thus produced are dense, compact and defect-free. A deeper characterization of these coatings will follow. To conclude, in a DEMO-oriented perspective, oxide coatings present interesting properties as multifunctional protective barriers, proving to be one of the few possible solutions for the material-related issues of all the breeding blanket concepts currently considered.
8:00 PM - EN17.09.16
Experimental and Computational Study of Lattice Anharmonicity Effects in Oxide Nuclear Fuels
Zilong Hua1,Amey Khanolkar1,Marat Khafizov2,Yuzhou Wang2,Chris Marianetti3,Lyuwen Fu3,David Hurley1
Idaho National Laboratory1,The Ohio State University2,Columbia University3
Show AbstractThermal transport in nuclear fuels under irradiation is directly linked to reactor efficiency and safety. An insight into the fundamental thermal transport properties of 5f-electron materials under extreme irradiation environments is central to the development of advanced nuclear fuels. Investigating the effects of lattice anharmonicity can provide a deeper understanding of the thermal properties such as thermal conductivity, heat capacity and thermal expansion, and be used to accurately model thermal transport. Here, we present the latest study on lattice anharmonicity effects in advanced oxide fuels. Time-domain Brillouin Scattering (TDBS) has been used to generate and detect lattice vibrations associated with the propagation of quasi-longitudinal and quasi-shear elastic phonon modes along the (311), (111), and (100) directions in single crystal thorium oxide (ThO2). Measurements of the temperature-dependent change in phonon velocities have been performed at various temperatures ranging from 77 up to 350 K to experimentally obtain the Gruneisen parameter. The value is then compared with those obtained from first-principles computation using density functional theory (DFT), and more related simulation and experimental work (i.e., Boltzmann Transport Equation (BTE) modeling, inelastic neutron and x-ray scattering measurements, and the characterization of thermal conductivity, heat capacity, and thermal expansion) are underway. This study is supported by Thermal Energy Transport under Irradiation (TETI), a recently awarded Energy Frontier Research Center. The mission of the center is to provide foundational work to understand the role of 5f electrons on phonon and electron structure of related materials and how irradiation defects influence electron and phonon transport and corresponding thermal conduction.
8:00 PM - EN17.09.17
Combined Molecular Dynamics and Sol-Gel Synthesis to Investigate the Properties of the Pellet-Clad Bonding Layer
Dillon Frost1,2,Jessica Carolan-Veliscek2,Conor Galvin1,Edward Obbard1,Michael Cooper3,Patrick Burr1
UNSW1,Australian Nuclear Science and Technology Organisation2,Los Alamos National Laboratory3
Show AbstractThe typical burnup of currently operating nuclear reactors is around 47 GWd/MTU, which coincides with the onset of permanent bonding between UO2 fuels and zirconium alloy claddings in water cooled reactors. The bonding between pellet and cladding occurs as a result of the oxidation of Zr-alloy cladding and subsequent bonding with the UO2 forming a solid solution of (U,Zr)O2. The impact of this pellet-clad bonded layer increases with increasing burnup. In order to extract more energy from nuclear fuels and increase time between refuelling, operators are increasing burnups beyond 50 GWd/MTU. This research looks at the impact of urania-zirconia mixed oxides [(U,Zr)O2] on nuclear fuels at high burnups.
The following thermophysical properties of (Ux,Zr1-x)O2 were investigated as a function of composition and temperature: lattice parameter, thermal expansion, heat capacity and thermal conductivity. A combination of molecular dynamics and experimental methods were used to provide a means of prediction and validation. Mixtures of (Ux,Zr1-x)O2, where x = 0.25, 0.50, 0.75 and 0.9, were produced using sol-gel synthesis to enhance the mixing of cations and achieve a homogeneous solid solution. The results indicate that by increasing the concentration of Zr in (U,Zr)O2 mixtures the thermal conductivity is degraded slightly at room temperature, becoming negligent at temperatures at and above reactor operating temperatures. However, the formation of t-ZrO2 on the inner surface of the cladding at moderate burnups (20 – 30 GWd/t) is a barrier to thermal transfer being that the thermal conductivity of t-ZrO2 is 50% lower than that of UO2. Additionally, the lower temperature access to superionic transitions due to oxygen disorder could provide some benefit under accident conditions due to the significant increase in heat capacity around 1500-2000 K. This research shows that the impact of (U, Zr)O2 at high burnups is positive in that it does not affect the thermal conductivity of UO2 significantly and, due to the permanent bonding of clad and fuel, could enhance thermal throughput given the thermal conductivity of the bonded layer is an order of magnitude greater than that of He.
8:00 PM - EN17.09.19
Measuring Retention and Erosion Properties of SiC and W-SiC Coatings Exposed to High-Flux Ion Irradiation in an RF Plasma Source
Gregory Sinclair1,Tyler Abrams2,Stefan Bringuier2,Dan Thomas2,Leo Holland2,Sean Gonderman2,Gokul Vasudevamurthy2,Jonathan Yu3,Russell Doerner3
Oak Ridge Associated Universities1,General Atomics2,University of California, San Diego3
Show AbstractSiC and W-SiC functionally-graded coatings have been fabricated and exposed to deuterium (D) plasma bombardment at different surface temperatures and ion impact energies in the PISCES-E radio frequency (RF) plasma source [1] to assess their viability as a plasma-facing material (PFM) in a magnetic confinement nuclear fusion device. In situ optical emission spectroscopy (OES) revealed that CD emission, which approximates C chemical erosion, increased with ion impact energy up to 80 eV. Chemical erosion of Si was not observed spectroscopically via the SiD molecular band. Post-mortem Auger electron spectroscopy (AES) measured a <15% increase in relative Si content within the first few hundred nm of the surface. Si enrichment was likely due to preferential erosion of C atoms, as indicated by the asymmetric spectroscopic emission mentioned above. Erosion due to physical sputtering may have also been higher for C than for Si, as shown in [2]. The amount of D retained in samples due to implantation increased by a factor of ~5 between 20 eV and 50 eV, likely due to the onset of ion-induced, extrinsic trap formation. Overall retention was 1.7-2.5× higher in SiC than in W exposed to identical plasma irradiation. The pure W surface of the W-SiC functionally-graded coatings survived plasma exposure at a fluence of 1025 m-2, as confirmed by AES measurements. Overall erosion of the W-SiC coating was around 2.3× higher than that of pure, crystalline W samples exposed to the same conditions. D retention in the W-SiC coatings exhibited combined trap behavior of both W and SiC samples, and was >10× higher than retention in SiC and W samples. High erosion and retention values may have been an artifact of the coating process, resulting in increased intrinsic trap density and reduced lattice displacement and surface binding energies.
SiC has been considered for use as a PFM in future magnetic fusion devices due to its excellent mechanical strength under neutron irradiation, high thermal strength, and low hydrogenic diffusivity. Tungsten is currently the leading PFM candidate, due to favorable thermal and mechanical properties and low erosion. However, concerns regarding its strength under intense neutron irradiation and its detrimental effect on plasma performance when eroded have led to research on other material candidates, such as SiC. Recent experiments in DIII-D showed that SiC exhibited lower levels of chemical sputtering than that of graphite (a common PFM) [3]. These results motivated further characterization of fundamental plasma-material properties in a dedicated plasma source. SiC samples (coating thickness = 100-200 µm) and W-SiC samples (coating thickness = 10 µm) were fabricated in-house via chemical vapor deposition (CVD) and RF magnetron sputtering, respectively, on ATJ graphite substrates. W-SiC coatings exhibited a stepwise concentration gradient of 100% W at the surface to 100% SiC at the coating-bulk interface. These samples, along with W and ATJ graphite reference samples, were exposed to D RF plasmas in PISCES-E at a flux of 5 × 1020 m-2 s-1, Te of ~ 3 eV, ne of 3.5 × 1016 m-3, and surface temperature of 500 K. In situ OES was used to track erosion products and was quantified via the SXB/DXB method. Overall erosion was measured using a microbalance with a resolution of 0.01 mg. AES was used to measure changes in surface composition due to plasma exposure. Post-mortem thermal desorption spectroscopy resolved the major traps present in the material and estimated overall retention. Qualifying new SiC-based materials in a dedicated plasma source may demonstrate a superior alternative for next-generation fusion reactor materials.
Work supported by General Atomics Internal Research & Development and the U.S. DOE under DE-FC02-04ER54698 and DE-FG02-07ER54912.
[1] G.R. Tynan et al., J. Vac. Sci. Technol. A. 15 (1997) 2885–2892.
[2] S. Bringuier et al., Nucl. Mater. Energy. 19 (2019) 1–6.
[3] T. Abrams et al., MRS Fall Meeting (2019).
8:00 PM - EN17.09.20
The Effect of Flux on the Irradiation-Induced Precipitation in AISI-316L—An In Situ TEM Study
Ítalo Oyarzabal1,2,Matheus Tunes2,Osmane Camara2,Emily Aradi2,Inamul Haq Mir2,Graeme Greaves2,Jonathan Hinks2,Paulo Fichtner1,Stephen Donnelly2
UFRGS1,University of Huddersfield2
Show AbstractStructural components of nuclear reactors must fulfil a series of mechanical and stability properties according to the specific range of experimental conditions they will be submitted during the operation of the reactor. Since the beginning of the construction of generation-II Pressurized Water Reactors (PWR’s) in the 1960’s, austenitic stainless steels from the AISI 300 series have been widely used as structural components for the fuel cladding, due to their high-temperature corrosion-resistance and mechanical strength. The variation of neutron flux along the core of a nuclear reactor presents a challenge to the materials used to build the various components of the core, considering that for a non-steady flux the effects on structural properties also varies. Wiedersich, Okamoto and Lam (WOL) proposed a model to specify irradiation conditions of dose rate and temperature where significant RIS should occurs, being limited by an increase in the thermal concentration of vacancies at high temperatures and recombination of vacancies with interstitials at higher dose rates. In this work, we report the formation of radiation-induced precipitation (RIP) on an AISI 316L steel under in-situ 325 keV Xe ion irradiation/implantation for three different fluxes (1, 2 and 4x1012 ions.cm-2.s-1) up to the same total dose within a transmission electron microscope (TEM) at MIAMI facilities at the University of Huddersfield, UK. All experiments were performed at 550 °C. Statistical analysis from bright-field images under phase contrast and selected area electron diffraction (SAED) were used to characterise the precipitates formed in the samples analysed. The nucleation and growth of second phase precipitation are observed to vary significantly according to the flux the sample is submitted. An overlapping of consecutive cascades at higher fluxes is proposed to explain the different behaviours for the evolution of RIP’s, and a small deviation on the lower limit of the WOL’s model is suggested.
Symposium Organizers
Maria Okuniewski, Purdue University
Chaitanya Deo, Georgia Institute of Technology
Maik Lang, University of Tennesee
Simon Middleburgh, Bangor University
EN17.10: Nuclear Materials—Ceramics and Composites
Session Chairs
Remi Dingreville
Arielle Miller
Wednesday AM, December 04, 2019
Sheraton, 3rd Floor, Hampton
8:15 AM - EN17.10.01
Radiation Damage in Ceramics—The View From Microscopy
Karl Whittle1
University of Liverpool1
Show AbstractIn the continued development of materials for use within a nuclear reactor, whether it be fission or fusion based, how a material behaves under the extremes of damage radiation is a key factor in its applicability. The impacts arising from radiation induced damage are generally damaging, ranging from local damage through to full amorphisation, or the formation of new phases both by equilibrium or non-equilibrium synthesis. Such changes often have macro-scale impacts, for example a reduction in thermal conductivity, or change in a ductility. However, they key length scale at which all changes can be considered is at the nano-scale, where damage can be visualised most easily.
Using electron microscopy as a key aid in determining the impacts from damage, a range of systems, primarily ceramic in nature, will be discussed. Such materials are key components in multiple next generation reactor technologies, whether it be fission or fusion, fuel or structural in application. Regardless of the application, there are common factors that influence a response to radiation induced damage, and the examples presented here will help link the common themes, whether they be compositional or structural in nature.
8:45 AM - EN17.10.02
Modelling of Crack Nucleation and Propagation in SiC/SiC Accident-Tolerant Fuel During Routine Operational Transients Using Peridynamics
Thomas Haynes1,Mark Wenman1
Imperial College London1
Show AbstractZirconium-based alloys have been successfully used for light water reactor fuel claddings since the 1960s. Despite their excellent performance during routine operation, during severe accidents such as the one which occurred at Fukushima Daiichi in 2011, they can undergo rapid exothermic oxidation. This releases hydrogen and supplies a great deal of heat, accelerating the accident. Over the last decade, there has been a concerted effort by the nuclear fission community to develop and implement accident-tolerant fuel designs. One design showing promise is SiC/SiC composite claddings. These consist of a β-phase fibre weave surrounded by β-phase SiC matrix. Despite their strength and improved oxidation performance, the proposed claddings have a number of drawbacks. Namely their irradiation swelling and low thermal conductivity. Their irradiation swelling saturates after a number of months in reactor and is greater on the colder outer surface than the warmer inner surface. The thermal conductivity lower than for zirconium alloys and is worsened by irradiation swelling.
Peridynamics is a developing non-local modelling technique which has the ability to model crack initiation, propagation and branching. Unlike finite element analysis, the technique uses an integral formulation to determine the forces upon material points. This means that crack patterns can be predicted without any a priori knowledge of the loading or crack pattern.
A peridynamic model for composite SiC/SiC cladding has been developed in the Abaqus finite element code. The material model was isotropic and considers matrix cracking and fibre pull-out. The thermal expansion, swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions. The swelling on the outer surface is greater than the inner surface due to the lower irradiation temperature, causing a tensile stress on the inside of the cladding. This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface. In models without fibres, cracks would propagate through the cladding. If fibres are modelled, matrix cracking will extend to a depth of around 20% through the cladding from the inner surface. If an inner monolith of SiC is additionally modelled, cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite. If an outer SiC monolith is modelled, fibre pull-out on the inner surface of the cladding was increased by just under 70%.
9:00 AM - EN17.10.03
Irradiation of Carbide Composites for Next Generation Nuclear Reactors
Daniel Rutland1
University of Liverpool1
Show AbstractFuture generations of nuclear reactors are required to operate at higher temperatures and under more extreme radiation conditions than the current generation, creating a need for novel materials for use in these environments. Silicon carbide (SiC) has a number of desirable characteristics for use in reactors but some properties can deteriorate under irradiation. A potential option to improve the radiation resistance is to combine SiC with other carbides such as Titanium carbide (TiC) and Zirconium carbide (ZrC).
To determine the properties of these samples, they can be exposed to high radiation environments and electron microscopy and diffraction techniques can be used for characterisation and analysis of the materials.
Initial samples of SiC/TiC composites have been produced and irradiated in several facilities, with bulk irradiation using 93 MeV Pb ions and in-situ irradiation using 1 MeV Kr ions. The in-situ TEM imaging indicates different responses rates to radiation damage in the TiC phases and the SiC phases. Diffraction of these areas indicates almost immediate amorphisation of the SiC phases, with TiC phases retaining distinct grain boundaries, which is supported by X-Ray Diffraction (XRD) of bulk irradiated samples. The results from these samples are used to guide the fabrication and characterisation of the next set of samples. These samples produced are SiC/ZrC and SiC/TiC composites, each with ratios of 30:70, 50:50, and 70:30. The raw powders are milled at low-speed in inert gas, before being sintered using spark plasma sintering (SPS) at 1600-1900 oC. The results of these samples will be discussed in this talk.
9:15 AM - EN17.10.04
Radiation Tolerance of Stabilized Alumina Coatings—An In Situ Irradiation Study
Matteo Vanazzi1,Davide Loiacono1,2,Wei Chen3,Meimei Li3,Marco G. Beghi2,Fabio Di Fonzo1
Istituto Italiano di Tecnologia1,Politecnico di Milano2,Argonne National Laboratory3
Show AbstractFuture generation nuclear reactors – either fission or fusion systems – present intrinsic benefits in terms of economics, safety and waste management. However, to allow the deployment of these technologies, innovative material strategies must be designed and implemented. Recently, coatings have earned great attention in the nuclear field as a reliable cost-efficient solution. Indeed, they could tackle major issues such as corrosion, erosion, hydrogen/tritium confinement, etc. In this framework, amorphous/nano-crystalline alumina coatings developed by the Istituto Italiano di Tecnologia (IIT) have been extensively characterized as corrosion-resistant anti-diffusion barriers. Al2O3 films have been tested under ion irradiation, proving high stability up to 150 dpa. This radiation tolerance has been related to the amorphousness of the pristine material, which evolves towards a crystalline structure under irradiation. To preserve these enabling mechanisms and improve the range of operation (in terms of life time and operation temperature), the amorphous matrix must be conserved as long as possible. In this work, the stabilization of alumina by doping is evaluated under different conditions. Firstly, pure and doped materials have been compared by thermal annealing, showing a significant increase in the crystallization temperature (from 650 °C to more than 800 °C) for the stabilized alumina. Then, irradiation tests have been performed with in situ TEM at the IVEM-Tandem facility, at ANL. Samples have been irradiated with different ions (namely, 350 KeV Kr and 500 KeV Au) up to 20 dpa. Experiments have been conducted at 600 °C - for the sake of consistency with previous ion irradiation tests - and at 800 °C, to observe the accelerating effect of irradiation on the thermally activated crystallization process. Albeit both the materials withstand the ion irradiation without delamination or structural failure, doped alumina retards the crystallization threshold even under irradiation. Moreover, at significant levels of radiation damage, the presence of doping stabilizes the low-energy metastable phases of alumina, conserving partially the pristine characteristics. This behavior, especially at the highest temperature, improve further the radiation tolerance of the material: while pure alumina suffers from voids formation after crystallization (a traditional phenome for the bulk oxide), no swelling appears in the doped counterpart. To conclude, in situ ion irradiation tests have been performed on stabilized nano-ceramic coatings. The presence of doping slows down the material evolution under irradiation, preserving the well-established properties of the pristine coating. The outcomes of these experiments represent a major step forward in the qualification process of coatings for nuclear reactors.
9:30 AM - EN17.10.05
Metal/Carbon Nanotube Composites Enhance Strength and Ductility, Even During Radiation Damage
Kang Pyo So1,Penghui Cao2,Young Hee Lee3,Michael Short1,Ju Li1
Massachusetts Institute of Technology1,University of California, Irvine2,Sungkyunkwan University3
Show AbstractThe accumulation of defects during irradiation leads to material property degradation modes such as embrittlement and swelling, eventually causing material failure. Effective and efficient removal of defects is of crucial importance to design radiation damage-tolerant materials. Here, by biasing defect migration pathways via carbon nanotube (CNT) infiltration, we present a greatly enhanced damage-tolerant Al-CNT composite with defect storage measured to be one order of magnitude lower than that in pure, irradiated Al. Furthermore, extreme-value statistics (largest size) of defect clusters are significantly changed in the presence of CNT. In situ ion irradiation transmission electron microscopy (TEM) experiments and atomistic simulations together reveal the dynamic evolution and convergent diffusion of radiation-induced defects to CNTs, facilitating defect recombination and enhancing radiation tolerance. The occurrence of CNT-biased defect convergent migration is tuned by the thermodynamic driving force of stress gradient in Al matrix due to the CNT phase transformation. This approach to controlling defect migration using 1D interface engineering creates a new opportunity to enhance the properties of nuclear materials.
9:45 AM - EN17.10.06
Investigation of Helium Precipitates in Ta(Ti)/Zr(Ti) Nanocomposites
Kelvin Xie1,Sisi Xiang1,Ian McCue1,Michael Demkowicz1
Texas A&M University1
Show AbstractMetal degradation due to implanted helium (He) is a concern in nuclear energy applications. We investigate the size and distribution of He bubbles in Ta(Ti)/Zr(Ti) composites using transmission electron microscopy. He implanted composites are compared to both the unimplanted material as well as the individual constituents of the composite. Nano-scale He bubbles form in different quantities in the two components of this composite as well as at the curved Ta(Ti)/Zr(Ti) interfaces within the materials. We discuss the consequences of our findings for the development of radiation resistant materials with complex, hierarchical microstructures.
EN17.11: Irradiation Simulations and Performance
Session Chairs
Thomas Haynes
Karl Whittle
Wednesday PM, December 04, 2019
Sheraton, 3rd Floor, Hampton
10:30 AM - EN17.11.01
Reduced-Order Atomistic Method for Simulating High Dose Irradiation in Metal
Remi Dingreville2,Elton Chen1,2,Chaitanya Deo1
Georgia Institute of Technology1,Sandia National Laboratories2
Show AbstractAtomistic modeling of irradiation damages through displacement cascades is deceptively nontrivial. Due to the high energy, high velocity nature of the atom collisions, individual cascade simulations can become very computational expensive and ill-suited for size and dose upscaling. In order to examine microstructural evolutions, and mechanical property changes due to defect accumulation, alternative methods of modeling radiation defect accumulation are needed. Originally developed for application in ceramic materials, the Frenkel Pair Accumulation (FPA) method generates point defect pairs by directly displacing atoms from its initial lattice site. The applicability of method is somewhat limited to metallic/dense materials, as it does not capture the important cascade process known as the thermal spike. The presence of the thermal spikes has shown to be influence both point defect clustering and sequential cascade overlaps. Instead, using FPA as the basis and incorporating a reduced-order approximation for thermal spike, a new method of modeling radiation damage is developed. By adopting the athermal recovery corrected (arc) formalisms, the new arc Damage Insertion (arc-DI) method is able to predict and replicate radiation events across a wide range of recoil energy. Using Cu and Nb as the case studies, arc-DI is verified against standard displacement cascades. Example applications for simulating high energy cascade fragmentation and large dose ion-bombardment are also provided for demonstration.
Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC., a wholly owned subsidiary of Honeywell International, Inc., for the U.S. Department of Energy National Nuclear Security Administration under contract DENA0003525.
10:45 AM - EN17.11.02
Exceptional Radiation Performance of Crystalline-Amorphous Nanocrystalline Metal
Miaomiao Jin1,2,Penghui Cao3,Michael Short1
Massachusetts Institute of Technology1,Idaho National Laboratory2,University of California Irvine3
Show AbstractNanostructured materials with amorphous intergranular films (AIFs) have demonstrated superior strength and ductility. The radiation performance is expected to be as high as grain boundaries, if not better, due to the large fraction of interfacial volumes, which can be efficient sinks for radiation-induced defects. Meanwhile, the concentrated solute atoms within the film can pin down interface migration. However, better understanding beyond qualitative speculation must be pursued towards the radiation behavior of such novel design for the potential deployment in nuclear applications. In this study, we examine the response of an AIF system (nanocrystalline Cu with Zr doped AIFs) under continuous irradiation with molecular dynamics (MD) simulations. The system behavior is evaluated from three perspectives: ion mixing, defect reduction, and interface evolution. We propose a diffusion model that well characterizes the cascade-driven mixing process, and the spread of Zr distribution scales linearly with the damage level. The exceptional radiation resistance of such design can be understood from three aspects: i) Zr mixing into the bulk enhances local defect annihilation, ii) Zr impedes radiation-enhanced grain growth by restraining AIFs from migration, and iii) Zr increases the interface stiffness to maintain AIF integrity. These findings suggest that the AIF-engineered system can be a promising candidate in applications where strong mechanical property, structure stability and self-healing capability of radiation damage are required.
11:00 AM - EN17.11.03
Quantitative Phase Field Model for Void Nucleation and Growth under Ion Irradiation
Rayaprolu Goutham Sreekar Annadanam1,Anter El-Azab1
Purdue University1
Show AbstractVoids develop in crystalline materials under energetic particle irradiation as in nuclear reactors. Understanding the underlying mechanisms of void nucleation and growth is of utmost importance as it leads to dimensional instability of the metallic materials. In the past decade, researchers have adopted phase field approach to study the phenomena of void evolution under irradiation. In the phase field approach, the boundary between the void and matrix is modeled as a diffuse interface, while it is sharp in reality. This consideration of diffuse interface requires the phase field models to asymptotically match with their sharp interface counterparts in the limit of vanishing interface thickness. The asymptotic analysis of the phase field models has enabled in obtaining a quantitative vacancy-based phase field model, however, it failed to match with the sharp interface model when included interstitials concurrently with vacancies. In this communication, we introduce a thermodynamically consistent, quantitative diffuse interface formulation of type C for void evolution under irradiation, that includes vacancy, interstitial and phase field variable as order parameters. Test cases are presented to validate our new phase field model.
11:15 AM - EN17.11.04
Investigation of Proton Radiation Damage Effects on the Tensile Strength of 3D Printed Acrylonitrile Butadiene Styrene
Arielle Miller1,Dharmaraj Raghavan1,Grant Warner1
Howard University1
Show AbstractPolymers are widely used in nuclear power plant (NPP) components and in the storage of radioactive materials. Radiation-induced polymer degradation is a known challenge to the lifetime of components. Additive manufacturing (AM) provides a unique capability to cost-effectively replace used parts to key NPP components, thereby extending the life of commercial NPP and reducing maintenance costs and reactor downtime. The radiation damage due to polymers by neutron irradiation is mainly due to secondary protons, recoiled from the polymers themselves. Therefore the study of the effects of proton irradiation of polymers informs the damage study of neutron radiation and is a viable simulation of neutron damage. The radiation damage of 3D printed polymers with gamma beams has already been studied in terms of changes to mechanical properties such as tensile strength and flexural strength. The study aims to obtain an understanding of the proton radiation impact on the mechanical performance of fused filament fabrication (FF) acrylonitrile butadiene styrene (FF-ABS). The study addresses the effect of proton radiation damage at 40 MeV and high-level proton doses through experimental evaluation of both mechanical performance and material characteristic changes. Furthermore, it considers the comparison of mechanical damage caused by protons versus that caused by Co-60 gamma-ray radiation. The key components of this research are (1) material characterization to investigate irradiation damage to FD-ABS parts caused by exposure to protons with energies up to 40 MeV and a maximum dose of 1 MGy; and (2) tensile experiments to investigate the ionizing radiation effects on the mechanical properties of the FD-ABS. Preliminary results from irradiation performed at Texas A&M University Cyclotron Insitute show a significant drop in tensile strength as the radiation dose increased. The molecular weight and decomposition temperatures decreased as the radiation dose increased.
11:30 AM - EN17.11.05
In Situ Cantilever Testing of Radiation Damage in Tungsten for Fusion Applications
James Darnbrough1,Markus Alfreider2,Daniel Kiener2,David Armstrong1
University of Oxford1,Montanuniversität Leoben2
Show AbstractTungsten has a high melting point and low sputtering rate, this makes it ideal for use in fusion reactor walls and divertors [1]. In this environment the material is exposed to temperatures up to 1000°C and 14.1MeV neutrons leading to 150dpa of damage in a lifetime [2]. For structural components (e.g. divertors) it is important to understand the effect that this extreme environment has on the mechanical properties throughout working lifetime. Previous work suggests that high irradiation doses can lead to embrittlement in tungsten [3].
To investigate this, Helium implantation of Tungsten with 2% Tantalum (WTa2%) (including the role of transmutants) was irradiated with He+ up to 3000appm at a temperature of 300°C [4]. The implantation was conducted using a range of energies (0.05-1.8MeV) and fluences (1-5x1015 ions/cm2) to produce a relatively flat implantation profile ~1.56μm deep.
This scale of damage therefore requires small scale characterisation to understand the mechanical properties that change during radiation damage. In this work we illustrate the use of nanoindentation and microscale cantilevers to find the changes in/to the elastic and plastic response of the material. We have developed a new geometry of cantilever in order to test material volumes that are removed from the edge of the sample. This methodology allows for observation of the elastic and plastic properties of irradiated material and opens up a wide range of possibilities for future work in terms of location specific characterisation.
[1]:- P. Norajitra et al Journal of Nuclear Materials 2007 367-370 pp 1416-1421
[2]:- N. Baluc et al Nuclear Fusion 2007 47 pp S696-S717
[3]:- N. Yoshida Journal of Nuclear Materials 1999 266-269 pp 197-206
[4]:- C. E. Beck et al Symposium HH – Advances in Materials for Nuclear Energy 2013, Vol 1514 pp. 99-104
11:45 AM - EN17.11.06
Understanding Relation between Microstructure and Oxidation Resistance of an Additively Manufactured Nickel-Based Superalloy
Guofeng Wang1,Grace V. de Leon Nope1,Weitao Shan1,Brian Gleeson1
University of Pittsburgh1
Show AbstractAdditive manufacturing (AM) refers to the technique of producing complex three-dimensional objects directly from a digital model through layer-by-layer computer-controlled deposition and consolidation of feedstock powder materials. The AM technique is emerging into a mainstream manufacturing process and has the potential to find widespread application in nuclear technologies. Inconel 625 is a Ni-Cr-based alloy widely used in nuclear reactors. In this study we investigated the relations between the microstructure and associated high-temperature oxidation properties of additively manufactured IN625. Using Laser Engineered Net Shaping (LENS) technology, we fabricated 1x1x1 cm cubes from argon-atomized IN625 powders. The microstructure of these LENS-processed IN625 consisted of primary dendrites and fine cellular grains. Characterization revealed a significant microsegregation of Nb and Mo at the boundaries of the fine grains due to local rapid solidification processes. Thermogravimetric analysis was then conducted to determine the intrinsic oxidation resistance of the LENS-processed IN625 samples exposed to dry air at 800 C and 1000 C for up to 100 h. The kinetics results indicated that the LENS-processed IN625 exhibited less resistance to oxidation as compared to wrought IN625. The results will be discussed in the context of Nb and Mo microsegregation to the grain boundaries detrimentally affecting the oxidation resistance of the LENS-processed IN625.
EN17.12: Nuclear Waste Materials
Session Chairs
Wednesday PM, December 04, 2019
Sheraton, 3rd Floor, Hampton
1:30 PM - EN17.12.01
We Need More Realistic Corrosion Tests to Provide Confidence in Safety Cases for Radwaste Disposal
William Lee1
Bangor University1
Show AbstractUnderstanding the behaviour of complex materials in complex environments over millennia is a challenge for both modellers and experimentalists but must happen to have confidence in the safety cases for permanent radioactive waste disposal. Current durability testing of ceramic materials (including glass, glass composite materials and cements) as hosts for radioactive wastes is based on overly simple materials and procedures including using powders or samples of defined size and shape, polished samples, distilled water, controlled flow rates, closely defined and constant temperatures, simple and inactive compositions (International Simple Glass, ISG) etc. While steps are being taken towards more realistic testing procedures to account for package corrosion products e.g. iron oxide from container, type of water in repository environments (salts), Engineered Barrier Systems (e.g. clay) and rock dissolution products etc. more is needed.
Recent work [1-2] has highlighted the need for improved standard testing procedures and modelling developments. In [1] the impact of partial water immersion on the International Simple Glass (ISG), a condition that is inevitable as a repository gradually becomes saturated with penetrating groundwater was examined. This revealed quite different degradation behaviour compared to standard tests in which the material is fully immersed. In particular, whilst in standard tests the system reaches a steady state with a very low alteration rate thanks to the formation of a protective gel layer, in partially-immersed tests this steady state could not be reached because of the continuous alteration from the condensate water film. The constant input of ions from the emerged part of the sample caused a supersaturation of the solution, which resulted in early precipitation of secondary crystalline phases. In [2] using highly radioactive 238Pu-doped and non-radioactive samples of borosilicate glass with chemical compositions and synthesis procedure similar to French standard SON68 glass showed dramatic differences in behaviour. Over 4 months the radioactive glass is fully covered by mechanically unstable alteration layer whereas the model glass remains virtually pristine. Such studies highlight the importance of understanding the evolution of (T, environment, activity) conditions with time and the inherent inhomogeneity of the wasteforms and how they change with time, and the need to work with active material.
This talk uses these, and other, data to highlight the need for improved durability testing of wasteforms.
[1] RK Chinnam, PCM Fossati and WE Lee, “Degradation of Partially Immersed Glass: A New Perspective,” J. Nuclear Mats. 503 56-65 (2018).
[2] BY Zubekhina, AA Shiryaev, BE Burakov, IE Vlasova, AA Averin, VO Yapaskurt and VG Petrov, “Chemical alteration of 238Pu-loaded borosilicate glass under saturated leaching conditions,” Radiochim Acta doi.org/10.1515/ract-2018-3097.
2:00 PM - EN17.12.02
Structure-Property Relationships of Copper Coating Materials for Canada’s Used Nuclear Fuel Containers
Jason Tam1,Bosco Yu1,2,Weiwei Li1,Jason Giallonardo3,Jane Howe1,Uwe Erb1
University of Toronto1,McMaster University2,Nuclear Waste Management Organization3
Show AbstractNuclear energy offers many benefits including stable electricity generation and low carbon output; however, one of the major challenges of nuclear energy is the management of the used fuel. In Canada, the Nuclear Waste Management Organization (NWMO) has proposed a deep geological repository (DGR) solution where used Canada Deuterium Uranium (CANDU) fuel bundles are packaged into a used fuel container (UFC), which is then placed at a reference depth of 500 m underground in a geologically stable rock formation and surrounded by bentonite clay. The UFC consists of a structural steel containment vessel which is copper coated on the exterior surface for corrosion protection. Two methods are used to copper coat the UFC. The first method is electrodeposition (ED) process which is used to coat the shell and hemi-spherical factory components in a conventional manufacturing environment. After UFC is loaded with used fuel, the vessel is welded shut and the UFC corrosion barrier is completed by applying cold spray (CS) copper coating to the weld closure zone under radiological conditions. The as-deposited CS copper coating exhibits high hardness and low ductility. In order to restore ductility, local annealing (1 hour at 350°C) with a heating band is performed.
Since the Cu on the UFC are applied using two different methods, it is important to develop a better understanding of the structure-property-processing-performance relationships of these coatings. In this study, we apply various electron microscopy techniques including electron backscattered diffraction (EBSD), focused ion beam (FIB) and transmission electron microscopy (TEM) to establish the relationships between the coating processing conditions (ED, CS, annealing of CS-Cu) and microstructure of the coatings at different regions including the ED-Cu and CS-Cu interfaces, as well as the Cu particle-particle interfaces in the CS-Cu coatings. These structural characterization results will provide a basis for the ongoing mechanical and corrosion testing programs seeking to advance knowledge on strength and corrosion resistance of these Cu materials for long term storage in the repository.
2:15 PM - EN17.12.03
Complexation/Speciation—New Insight Studies of the Secondary Phase Formation under Repository Conditions
Nieves Rodríguez-Villagra1,L.J. Bonales1,J. Cobos1
CIEMAT1
Show AbstractIn a deep geological repository scenario, oxidized uranium in aqueous systems will be stabilized as UO22+ (hexavalent uranium), as a consequence of tetravalent uranium oxidation by radiolytic byproducts. Uranyl cationic species (UO22+) in different complexation forms, such as uranyl oxyhydroxide hydrates are expected to be found at the whole pH range conditions. The importance of uranyl lies in its potential incorporation of trace radioelements onto secondary uranyl phases. In view of the difficulty of uranium chemistry in natural groundwater, it is worthy to improve speciation assessment techniques so as to characterize precipitated solid phases and understand chemical processes.
Recently Raman spectroscopy has been shown as powerful tool to analysis the speciation of various actinyl (UO22+,NpO2+ and PuO22+) and to determine distribution of those elements which are more likely to be stable in a near-field groundwater environment. Therefore, the aim of this work is to develop a quantitative procedure to follow uranyl speciation in aqueous solution under repository conditions, and understand the bonding, coordination and local surroundings of uranyl species, which will brings light to understand and predict the precipitated solid phases formed (solubility-limiting phases). Hydroxide and hydroxo-carbonate complexes of uranyl will be spectroscopically identified in aqueous solution in the absence and presence of carbonate. The proposed identification procedure of uranyl complexation will give a precise basis to assign better estimates of the predicted solid phase stoichiometry combining solubility analysis with speciation and precipitation of U(VI).
In this work, it is shown the use of Raman spectroscopy adapted to the empirical analysis of different nuclear applications for uranium concentrations higher than 0.05M at ambient atmosphere, i.e. as monitoring tool for UO22+ precipitation as a function of pH studying UO2(NO3)2x6H2O stability in aqueous solutions representative of groundwater, in particular at ionic strength I = 0.02 – 0.4 M and pH from 7 to 13.2 and to evaluate the effect of gamma radiation fields. Therefore, the obtained results will provided a complete picture of secondary phase formations, as a result of corrosion of spent nuclear fuel in a deep geological repository.
3:30 PM - EN17.12.04
Radiation Tolerant Ceramics for Nuclear Waste Immobilization—Structure and Stability of Zinc Substituted Hollandites with High Cesium Loading
Kyle Brinkman1
Clemson University1
Show AbstractThe radiation damage tolerance of nuclear waste forms is dependent on the materials resistance to defect formation and its ability to accommodate the structural distortions that arise from defect formation. This study illustrates how the radiation tolerance of hollandite can be improved thorough compositional control of cesium stoichiometry. The barium (Ba) to cesium (Cs) ratio was varied in the tunnel sites referred to as the A site of the hollandite structure. Zinc (Zn) was substituted for titanium (Ti) on the B site to achieve the targeted stoichiometry with a general formula of BaxCsyZnx+y/2Ti8-x-y/2O16(0<x<1.33; 0<y<1.33). The tunnel cross-section depended on the average A site cation radius while the tunnel length depended on the average B site cation radius. The enthalpies of formation from binary oxides of Zn doped hollandite measured using high temperature oxide melt solution calorimetry were strongly negative, indicating thermodynamic stability with respect to their parent oxides. The formation enthalpies became more negative, indicating hollandite formation is more energetically favorable, when Cs was substituted for Ba across the range of Zn doped compositions investigated in this study. This hollandite series was exposed to heavy ion (Kr2+) irradiation at 27 °C, 100 °C, 200 °C and 300 °C followed by characterization with grazing incidence X-ray diffraction , transmission electron microscopy , and aqueous leaching tests.. The radiation tolerance increased at elevated temperatures with a critical amorphization temperature between 200 °C and 300 °C. Elemental leaching decreased with increasing cesium content. Irradiated samples exhibited twice the fraction cesium release of pristine samples.
4:00 PM - EN17.12.05
Ion Irradiation-Induced Volume Swelling and Microcracking in Multiphase Glass Ceramic and Crystalline Ceramic Nuclear Waste Forms
Ming Tang1,2
Los Alamos National Laboratory1,Clemson University2
Show AbstractCrystalline ceramics and glass ceramics are candidate host materials for immobilizing alkaline/alkaline earth + lanthanide + transition metal fission product waste streams from nuclear fuel reprocessing. The major phases in these multiphase borosilicate glass ceramics are powellite, oxyapatite, pollucite, celsian, and durable residual glass phases. Al2O3 and TiO2 were combined with these waste components to produce multiphase crystalline ceramics containing hollandite-type phases, perovskites, zirconolite/pyrochlores and other minor metal titanate phases. These alternative waste form materials offer increased solubility of troublesome components in crystalline phases compared to glass. This, in turn, leads to increased waste loading. During the long term storage, radiation stability and chemical durability always are most important concerns for nuclear waste forms. The focus of this presentation is the radiation-induced volume swelling and microcracking. These radiation damage effects would affect chemical durability of nuclear waste forms.
In this study, selected glass ceramic and crystalline ceramic samples are exposed to various ion beam irradiations, which are used to simulate self-radiations in a waste form. Several individual crystalline phases and pure glass are fabricated and tested under the same ion irradiations. Specifically, light ion (Helium) irradiation is used to simulate alpha particles and ionization process; and heavy ion (Krypton) irradiation is used to simulate alpha recoil and produce displacement damage. Ion irradiation-induced microstructural modifications, volume swelling and microcracking are examined using X-ray diffraction, transmission electron microscopy, scanning electron microscopy, atomic force microscopy, and other characterization methods. Our preliminary results show that (1) similar radiation damage responses from single crystalline phases and corresponding crystalline phases in multiphase samples; (2) different crystalline phases in these multiphase waste forms exhibit different radiation-induced volume swelling and microcracks formation under the same ion irradiation damage environment, and different ion species radiations induce different volume swelling and microcracks in the same crystalline phase; (3) helium irradiation induced volume swelling and microcracks in glass and some crystalline phases are similar to the previous results of corresponding glass and crystallines irradiated by neutron or doped by actinide.
4:30 PM - EN17.12.07
Physics-Based Mesoscale Models of Ion Exchange in Hierarchical Materials
Yulan Li1,Shenyang Hu1,Chuck Henager1,Theodore Besmann2,Agnes Grandjean3,4,Hans-Conrad Zur Loye2
Pacific Northwest National Lab1,University of South Carolina2,Univ Montpellier3,CEA4
Show AbstractThe performance of advanced nuclear wasteform materials containing radionuclide absorbing nanoparticles such as zeolite, hexacyanoferrate (HCF), and salt-inclusion materials (SIMs) depends on hierarchical microstructures as well as thermodynamic and kinetics properties of ion exchange processes. For example, the wasteform morphology including particle size and particle aspect ratio, anisotropic mobility, and electrochemical potential affect the uptake kinetics of radionuclides. A physics-based mesoscale model of ion exchange in hierarchical materials is presented that takes into account the effects on ion exchange processes of inhomogeneous microstructures, anisotropic and inhomogeneous diffusivity of charged particles, electrochemical potentials, and operational conditions. Wasteforms consisting of zeolite or SIMs particles are taken as model systems. The thermodynamic and kinetic properties of diffusive ions are assessed using atomistic simulations and thermodynamic calculations. Simulation results showed that anisotropic mobility and larger aspect ratios of particles dramatically decrease ion exchange kinetics if the migration mobility is faster along the largest dimension anisotropic particles, such as needle-shaped particles. Electric fields associated with inhomogeneous ion distributions is an additional driving force that typically increases ion exchange kinetics depending on particle aggregation and ion mobility. The capability of the model for investigating the physics and mechanisms behind ion exchange in complicated microstructures and for predicting the effects of microstructure on ion exchange kinetics will be presented.
4:45 PM - EN17.12.08
Progress in Flash Sintering of UO2
Erofili Kardoulaki1,Darrin Byler1,Ken McClellan1
Los Alamos National Laboratory1
Show AbstractEarly studies on the onset of flash sintering (FS) for UO2 have identified that flash is induced by a critical field, at a critical onset time (incubation time) and they are both shown to be dependent on temperature and O/M. UO2+x has been shown to flash at lower critical fields, compared to UO2.00, and a bifurcation stability model has been used to model the Joule heating effects [1] taking place during flash onset with reasonable agreement with experimental data [2]. Densification of stoichiometric and hyper-stoichiometric UO2 has been achieved via FS and has resulted in 81% and 92% theoretical density (TD) pellets, respectively [3]. Still, the resulting microstructures were significantly cracked and featured porosity. The latest studies on FS of UO2 have concentrated on producing high density pellets with structural integrity and uniform microstructure. FS using the current control method has been implemented to minimize thermal shock and achieve control in the resultant microstructures. The current control method has been shown to produce more structurally sound pellets, although additional difficulties with FS of UO2 result from atmosphere and subsequently stoichiometry control of the pellet. More specifically, during FS of UO2+x using Direct Current (DC), an O/M gradient can develop due to the concentration of O2 close to the anode, resulting in hyper-stoichiometric material in the vicinity which can sinter at a faster rate. This produces pellets with differential densification along their length. To overcome this issue, a pseudo-Alternating Current (pseudo-AC) has been used and it has been shown to produce pellets with slightly higher densities and more homogeneous macroscopic characteristics. Finally, the pseudo-AC and current controlled methods have been applied to FS of doped UO2 and the results are compared to doped UO2 that has been sintered conventionally.
[1] J. G. P. da Silva, et al., “A dynamic bifurcation criterion for thermal runaway during the flash sintering of ceramics,” J. Eur. Ceram. Soc., vol. 36, no. 5, pp. 1261–1267, 2016.
[2] A. M. Raftery, J. G. P. da Silva, et al., “Onset conditions for flash sintering of UO2,” J. Nucl. Mater., vol. 493, pp. 264–270, 2017.
[3] J. A. Valdez, D. D. Byler, E. Kardoulaki, et al., “Flash sintering of stoichiometric and hyper-stoichiometric urania,” J. Nucl. Mater., vol. 505, pp. 85–93, 2018.
[4] D. Sprouster, E. Kardoulaki, et al., “In situ X-ray Characterization of Uranium Dioxide During Flash Sintering,” Materialia, vol. 2, pp. 176-182, 2018.
Symposium Organizers
Maria Okuniewski, Purdue University
Chaitanya Deo, Georgia Institute of Technology
Maik Lang, University of Tennesee
Simon Middleburgh, Bangor University
EN17.13: Materials for Nuclear Detection and Forensics
Session Chairs
Thursday AM, December 05, 2019
Sheraton, 3rd Floor, Hampton
8:30 AM - EN17.13.01
Crystal Growth and Scintillation Properties of Ternary Halides for Gamma Spectroscopy
Matthew Loyd1,Mariya Zhuravleva1
The University of Tennessee, Knoxville1
Show AbstractIn this report, we will review the crystal growth and scintillation properties of simple and mixed cation and mixed anion ternary halides that belong to the ABX3, AB2X5, A4BX6 families (A – one or more alkaline metals, B – one or more alkaline earth metals, X – one or a mixture of chlorine, bromine or iodine). They are of interest for scintillation applications due to their relatively high density, congruent melting, large band gaps, and availability of a divalent lattice site to accommodate a Eu2+ luminescence activator. They have attractive spectroscopic performance that is competitive with the industry standard, NaI:Tl, or state-of-the-art SrI2:Eu and LaBr3:Ce scintillators, and therefore are suitable for use in portable radiation detectors for radio-isotope identification. Details of compositional uniformity, growth defects, and gamma-ray spectroscopic performance as a function of Eu concentration, crystal volume and growth parameters are presented to provide understanding of the best achievable performance. Particular attention is paid to thermal and structural properties to design the crystal growth parameters via the vertical Bridgman method. Additionally, the potential for scale-up is evaluated by simultaneous growth of multiple large size crystals (1” – 1.5” in diameter) using the Multi-Ampoule Growth Station.
9:00 AM - EN17.13.02
Robust Perovskite Single Crystal Devices for Efficient Gamma-Ray Detection
Jeremy Tisdale1,Michael Yoho1,Shreetu Shrestha1,Kasun Fernando1,Sergei Tretiak1,Duc Vo1,Wanyi Nie1
Los Alamos National Laboratory1
Show AbstractGamma-ray detection and spectroscopy is the quantitative analysis of gamma energy spectra, and is of critical importance in many applications, such as nuclear safeguards, nuclear forensics, and many more. Recently, single-crystalline, hybrid (and inorganic) perovskites have been proposed as new semiconducting materials for gamma-ray spectroscopy. In just a few short years, early reports have shown promising results for perovskite-based semiconductor gamma detectors. Some of the most impactful results in this new application for perovskite materials include: CsPbBr3 showing gamma-ray spectra with a resolution of 3.8% at room temperature for a 662 keV 137Cs source at room temperature,1 MAPbI3 showing gamma-ray spectra with a resolution of 6.8% at a lower temperature of 2 °C for a 122 keV 57Co source,2 and MAPbBr2.94Cl0.06 showing gamma-ray spectra with a typical resolution of 12% for a 662 keV 137Cs at room temperature.3 Although first reports for this class of materials have shown significant results with high-resolution gamma detectors, many new issues have arisen for this new application of radiation sensing.
High electric field device stability and ion migration are deeply explored to understand the effects of high bias applications in hybrid perovskites and ways to suppress/eliminate the negative effects towards high-resolution, room temperature gamma-spectroscopy. In a Cr/MAPbBr2.85Cl0.15/Cr device, low electric fields (≤25 V/mm) pulses with extremely low rise times (average of 30 to 65 µs). In order to increase the rise time, the device needs to be operated at higher electric fields, (≥ 50 V/mm) which increases the rise time to an average of 15 µs. However, at high electric fields, the signals become noisy in roughly 10 seconds to 1 minute. We directly probed the cause of the noise using long-term high voltage biasing, while studying the hysteretic behavior via IV characteristics. A new Cr/MAPbBr2.85Cl0.15/Cr exhibits pure Ohmic behavior with negligible hysteresis between -200 to +200 V (E = 100 V/mm). However, after biasing the device for only 10 minutes at +200 V, a large hysteresis loop is observed. After letting the device rest in the dark for two days, the hysteresis loop becomes even larger, rather than restoring to its original properties, showing that the interfacial damage caused is non-reversible. Using the same electrodes in the device architecture allows us to use a technique termed voltage cycling to avoid permanent interfacial damage while operating the detectors. The signals from positive and negative polarity are roughly the same in the metal/semiconductor/metal device. Therefore, when noise appears, we are able to switch polarity to eliminate the noise and continue counting for gamma spectra. Cooling the detector to temperatures of -30 °C further increases the speed of the detector pulses by about 30%. However, the standard deviation in the pulse heights and rise times are still quite large, resulting in low resolution for the detectors with about 30-35% resolution for 59.6 keV γ-rays and 20-25% resolution for 662 keV γ-rays. Further development in hybrid (or non-hybrid) perovskite materials, such as CsPbBr3 and MAPbI3, should prove beneficial, as they are more stable under high electric fields required to achieve high resolution γ-ray spectroscopy. Also, further development in areas such as interfacial engineering and pixelated detector designs will be the next required steps in investigating the improvement for single crystalline, perovskite radiation detectors.
References
1. He, Y., et al., High spectral resolution of gamma-rays at room temperature by perovskite CsPbBr3 single crystals. Nature Communications, 2018. 9(1), 1609.
2. He, Y., et al., Resolving the Energy of γ-Ray Photons with MAPbI3 Single Crystals. ACS Photonics, 2018. 5(10), 4132-4138.
3. Wei, H., et al., Dopant compensation in alloyed CH3NH3PbBr3−xClx perovskite single crystals for gamma-ray spectroscopy. Nature Materials, 2017. 16, 826.
9:15 AM - EN17.13.03
Passive Radiofrequency Dosimeter Tag Based on Flexible Radiation-Sensitive Oxide Field-Effect Transistor
Tobias Cramer1,Ilaria Fratelli1,Pedro Barquinha2,Rodrigo Martins2,Elvira Fortunato2,Beatrice Fraboni1
University of Bologna1,Universidade Nova de Lisboa2
Show AbstractDistributed ionizing radiation dosimetry is crucial in diverse security areas with significant environmental and human impacts such as nuclear waste management, radiotherapy, or radioprotection devices. We present a fast, real-time dosimetry detection system based on flexible oxide thin-film transistors that show a quantitative shift in threshold voltage of up to 3.4 V/gray upon exposure to ionizing radiation. The transistors use indium-gallium-zinc-oxide as a semiconductor and amultilayer dielectric based on silicon oxide and tantalum oxide. [1] Our measurements demonstrate that the threshold voltage shift is caused by the accumulation of positive ionization charge in the dielectric layer due to high-energy photon absorption in the high-Z dielectric. The high mobility combined with a steep subthreshold slope of the transistor allows for fast, reliable, and ultralow-power readout of the deposited radiation dose. The order-of-magnitude variation in transistor channel impedance upon exposure to radiation makes it possible to use a low-cost, passive radiofrequency identification sensor tag for its readout. In this way, we demonstrate a passive, programmable, wireless sensor that reports in real time the excess of critical radiation doses. [2]
[1] T. Cramer, A. Sacchetti, M. T. Lobato, P. Barquinha, V. Fischer, E. Fortunato, R. Martins and B. Fraboni,“Radiation-Tolerant Flexible Large-Area Electronics Based on Oxide Semiconductors,” Adv.Electron.Mater., vol. 2, no. 7, pp. 1–8, 2016.
[2] T. Cramer, I. Fratelli, P. Barquinha, A. Santa, C. Fernandes, F. D’Annunzio, C. Loussert, R. Martins, E. Fortunato, B. Fraboni, “Passive radiofrequency X-ray dosimeter tag based on flexible radiation sensitive oxide field effect transistor”, Sci.Adv., vol. 4, eaat1825, 2018.
9:30 AM - EN17.13.04
Thermal Diffusivity Degradation and Defect Density Prediction in Self-Ion Implanted Tungsten Using Transient Grating Spectroscopy
Mohamed Abdallah Reza1,Hongbing Yu1,Kenichiro Mizohata2,Felix Hofmann1
University of Oxford1,University of Helsinki2
Show AbstractWe study the degradation of thermal diffusivity in self-ion implanted tungsten, and utilise this degradation to predict point defect densities. The defect density predictions are consistent with previous transmission electron microscopy (TEM) and molecular dynamics (MD) studies of self-ion implanted tungsten.
Tungsten, due to its high melting point, high thermal diffusivity and low sputtering yield, is the prime candidate for plasma facing armour components in future fusion reactors. Armour components in the so-called divertor, the exhaust of the fusion reactor, will receive the highest heat and radiation loads in the reactor. The damage created in the tungsten matrix as a result of intense neutron irradiation and plasma exposure decreases its thermal diffusivity, reducing its heat removal capability. Self-ion implantation is a convenient, low cost proxy to neutron irradiation to study the displacement damage created by neutrons. It efficiently singles out the displacement damage from other effects such as transmutation. However self-ion implantation even at energies as high as 20 MeV create damage layers that are only a few microns thick. The transient grating spectroscopy method used in this study enables us to measure the thermal diffusivity of these thin layers with high accuracy. In this study, a simple kinetic theory model is used to predict the point defect densities from the measured changes in thermal diffusivity as a function of ion dose. These predictions are compared to dislocation loop and point defect densities observed in TEM and predicted by MD.
20 MeV self-ion implanted tungsten, over a dose range of 0.0001 – 10 dpa, shows a significant degradation in thermal diffusivity above doses of 0.001 dpa. This degradation progresses up to 0.01 dpa, beyond which the decrease in thermal diffusivity saturates. Full saturation is achieved by 0.1 dpa, and the thermal diffusivity ceases to deteriorate with additional implantation. The reduction in room temperature thermal diffusivity at this damage level is 55% of the value of pristine tungsten.
The extracted point defect densities from the TGS data are about 2 orders higher than anticipated when considering solely defects visible in TEM. By combining TEM observations with predictions from MD, the density of point defects associated with defect clusters too small to be seen by TEM can be estimated. Once these “invisible” defects are accounted for, the total point defect concentration agrees to within an order of magnitude with that inferred from the TGS measurements of thermal diffusivity.
The reduction in thermal diffusivity observed in this study provides a much more complete picture of the degradation of thermal diffusivity with increasing damage dose. It also suggests that the diffusivity degradation due to the displacement damage from neutron irradiation will saturate beyond 0.1 dpa, which is an important result for predicting the in-service performance of armour components in future fusion reactors. The defect prediction results demonstrate the novel capability of TGS in quantitatively determining defect densities in irradiation-damaged materials. It also shows that the finely distributed smaller defects affect material properties such as thermal diffusivity to a larger extent than the bigger dislocation loops that are visible to TEM.
9:45 AM - EN17.13.05
Processing-Structure Related Performance of Irradiated Relaxor-Ferroelectric Thin Films
Evelyn Chin1,Cory Cress2,Nazanin Bassiri-Gharb1
Georgia Institute of Technology1,U.S. Naval Research Laboratory2
Show AbstractFerroelectric (FE) materials show a spontaneous polarization, reversibly switchable with an electric field, in addition to large dielectric, pyroelectric and piezoelectric response. Such properties make them attractive for fulfilling multiple functionalities in MEMS devices as sensors, actuators, and energy harvesting units. Traditionally, lead zirconate titanate (Pb[Zr1-xTix]O3, PZT) has been the most used ferroelectric in piezoelectric MEMS. However, relaxor-FE bulk single crystals, such as (1-x)Pb(Mg1/3Nb2/3)O3-xPbTiO3, (1-x)PMN-xPT, exhibit even larger electromechanical response than ceramic PZT, when cut and poled along the (001) direction with compositions on the rhombohedral side of the morphotropic phase boundary (x~0.32).
Lead-based materials are often radiation-hard, and numerous studies have shown PZT’s exceptional capabilities of retaining functional response after radiation exposure. The radiation hardness enables applications in high radiation exposure environments (e.g., aerospace, medical physics, x-ray/high energy source measurement tools, continuous monitoring of nuclear power plant applications, radiation disaster relief, etc.). However, PZT films still suffer from response degradation, especially for radiation doses at, or superior to 5 Mrad(Si). The high B-site chemical heterogeneity and disorder of PMN-PT is expected to result in a more radiation-hard material than PZT. Such perceived advantages are offset by challenges in processing pore- and crack-free PMN-PT thin films. Here we report on the radiation hardness of 0.7PMN-0.3PT thin films with varying microstructures, as modified through processing parameters.
All films were deposited on platinized Si substrates through chemical solution deposition (CSD). A 0.3M precursor solution of PMN-PT was prepared through a 2-methoxyethanol route and deposited onto the substrate via spin coating. A PbTiO3 seed layer was used to induce (100) crystallographic orientation in PMN-PT films. Heat treatment conditions were modified in order to control porosity and grain morphology. Specifically, the samples were pyrolyzed at 250 to 430 °C for 1 to 5 minutes, and annealed at temperatures of 700 to 800 °C for 1 or 2 minutes. The deposition and heat treatment process were repeated in order to obtain final film thicknesses up to 500 to 600 nm. All samples showed strong (100) perovskite orientation with Lotgering factors > 80%, as evidenced by X-ray diffraction patterns. The samples were irradiated using a 60Co source with ionization doses ranging from 0.2 to 10 Mrad(Si). A zero-dose sample was used for aging reference and background radiation control. Dielectric, polarization, and piezoelectric responses were conducted using sputtered Pt top electrodes, before and after irradiation.
In general, higher crystallization temperatures led to larger grains and columnar grain growth. All PMN-PT samples showed degradation in dielectric and saturated polarization within 5%, and in piezoelectric coefficients within 15%, within the radiation dose range used. Some films showed improvement in response up to 5% at lower doses, possibly due to annihilation of defects through low radiation exposure. Greater radiation hardness and lower functional response was observed in films with smaller, stacked grains, compared to films with larger, columnar grains. The influence of grain microstructure and porosity on both functional response and radiation-hardness, and a phenomenological model to quantify defect-defect interactions in irradiated functional materials will be discussed.
10:30 AM - EN17.13.06
Quantifying Morphological Features of Uranium Oxides for Nuclear Forensics
Luther McDonald1
The University of Utah1
Show AbstractIn the past five years, it was only hypothesized that morphological features could be used as signatures for nuclear forensics. The physicochemical connection between the processing parameters such as precipitation conditions, calcination temperature, and the composition of the starting material was generally thought to affect the morphology of the resulting material. However, quantifying these features was the limitation that prevented this signature from being utilized. In the past two years, we have published multiple papers demonstrating quantitative morphological analysis of UO3, U3O8, and UO2. To move from qualitative morphological analysis to quantitative analysis, the SEM images were processed using the Morphological Analysis of MAterials (MAMA) particle segmentation software. Following particle segmentation, the MAMA software was used to calculate morphological particle attributes for each of the segments. We generally segment enough images to acquire several hundred particles for statistical analysis. Our initial studies focused on synthetic parameters from the uranyl peroxide route in which we showed quantitative differences in the particle morphology of Am-UO3 and U3O8 based on the calcination temperature.
We have also shown that incorporation of impurities with the U-oxide will alter the final product morphology. For example, four impurities: calcium, magnesium, vanadium, and zirconium were separately introduced to pure uranyl nitrate hexahydrate from stock solutions. Following the addition of the impurities, the solutions were precipitated using hydrogen peroxide. Incorporation of the impurity was quantified using inductively coupled plasma mass spectrometry (ICP-MS). The uranyl peroxide precipitate was then calcined to U3O8 at 800°C. ICP-MS was used again, to evaluate the incorporation of the impurity in the calcined uranium oxide. SEM was used to investigate morphological changes, and SEM-Energy dispersive spectroscopy (EDS) was used to identify any concentrated regions of the impurity in the SEM images. Results will be discussed highlighting the potential of using particle morphology for nuclear forensics, but also highlighting the challenges that need to be overcome for this to be a viable signature of processing history.
11:00 AM - EN17.13.07
Non-Destructively Uncovering Aging and Irradiation Relationships for Nuclear Material Health Assessment and Forensics
Michael Short1,Cody Dennett1,Saleem Aldajani1,Benjamin Dacus1,Caitlin Huotilainen2,Ulla Ehrnsten2,M. Grace Burke3,Kudzanai Mukihawa3,Ihor Radchenko4,Kai Chen4,Ziv Ungarish5,6,Michael Aizenshtein6,Eyal Yahel5,Pål Efsing7,Thak Sang Byun8,Joe Wall9
Massachusetts Institute of Technology1,VTT Technical Research Centre2,University of Manchester3,Xi’an Jiaotong University4,Nuclear Research Center Negev (NRCN)5,Ben-Gurion University of the Negev6,Vattenfall AB7,Pacific Northwest National Laboratory8,Electric Power Research Institute (EPRI)9
Show AbstractNuclear materials age in service, with their properties changing as a combined result of how they were produced and to what they were exposed. This aging, which can be due to a combination of thermal and irradiation origins, sometimes results in degraded performance which can lead to failure. It also often leaves fingerprints, measurable either directly or indirectly. In this talk we will explore the use of transient grating spectroscopy (TGS) as a method to both assess the health of key light water reactor (LWR) components due to spinodal decomposition, short range ordering, and phase precipitation, and to reconstruct their irradiation history. Changes in thermal, elastic, and acoustic material properties are indirectly linked to the quantities of interest, such as Charpy impact energy via surface acoustic wave peak splitting for structural materials, and irradiation dose history to reconstruct reactor usage, particularly for additively manufactured materials. This wide variety of examples highlights the vast, unexplored space where non-destructive, indirect techinques can far more rapidly uncover new science and assist industry in nuclear materials forensic and health assessment issues.
11:30 AM - EN17.13.08
Thermodynamics and Electronic Structure of Actinide Based Metal-Organic Frameworks from Density Functional Theory Calculations
Shubham Pandey1,Zhilin Jia1,Brian Demaske1,2,Otega Ejegbavwo3,Natalia Shustova3,Wahyu Setyawan4,Chuck Henager4,Simon Phillpot1
University of Florida1,Sandia National Laboratories2,University of South Carolina3,Pacific Northwest National Laboratory4
Show AbstractMetal-Organic Frameworks (MOFs) are emerging as a novel class of hybrid materials for the purpose of radionuclide sequestration. Significant research has been reported on MOFs for applications such as gas storage, sensing and heterogeneous catalysis; however, radionuclide incorporated MOFs are relatively unexplored. The unique and desirable properties of MOF structures like high porosity, modularity, and synthetic diversity make them attractive candidate materials to contain the radionuclides present in the nuclear wastes. Density Functional Theory (DFT) calculations were used to determine the favorability of ion-exchange (at the metal node) of relevant radionuclides in various parent MOF clusters. A range of DFT methods, including various flavors of exchange-correlation functionals, DFT+U, relativistic effects, and magnetic effects were employed to establish robustness in results. The electronic structures of MOFs are also determined and are correlated with experimental data.
11:45 AM - EN17.13.09
Nanoscale Cross-Plane Thermal Transport and Structural Characterization of Swift Heavy Ion Irradiated LiF Crystals
Ainur Koshkinbayeva1,Azat Abdullaev1,Zhanatay Nurekeyev1,Vladimir Skuratov2,3,4,Zhandos Utegulov1
Nazarbayev University1,Flerov Laboratory of Nuclear Reactions, Joint Institute for Nuclear Research2,National Research Nuclear University MEPhI3,Dubna State University4
Show AbstractThermal transport in insulators is of great interest in high temperature nuclear reactors application. External irradiation by swift heavy ions (SHIs) such as Bi at hundreds of MeV energy is used to simulate the fission product impact in nuclear and model materials like LiF with well-known structural response to dense excitation effects [1, 2].
We report the results of color centers formation analysis in bulk lithium fluoride crystals irradiated at room temperature by 98 MeV, 244 MeV, 465 MeV and 710 MeV Bi ions at fluence 1011 cm-2 and 710 MeV ions at fluencies ranging from 1010 to 1013 cm-2 using optical absorption (OA) and photoluminescence (PL) spectroscopy.
The irradiation of LiF samples was carried out at the U-400 FLNR JINR cyclotron (Dubna, Russia). The proliferation of ion-induced F (245 nm) and F2 (445 nm) color centers were studied by OA spectroscopy. As the fluence of irradiation ions grows, the integrated absorption intensity of F-peak gradually decays, whereas that of F2-peak broadens and increases, which indicates that color centers tend to aggregate. The corresponding defects densities were estimated using Smakula-Dexter formula [2].
Since F2 and F3 color center related bands in absorption spectra overlap, the complementary technique, PL spectroscopy, was performed. Specifically, the F2 and F3 color centers in irradiated LiF samples were distinguished by their characteristic peaks observed at around 530 and 670 nm, respectively. For the fluencies 1010 - 1013 cm-2, the integrated fluorescent intensity of F2 appears to dominate in the spectrum for the lower ion doses and drops to the level comparable to that of F3 as the ion dose increases.
Cross-plane near-surface thermal conductivity of SHI irradiated LiF crystals was characterized by picosecond time-domain thermo-reflectance (TDTR) in the thermally probed region extending down to 290 nm, where the electronic stopping / ionization processes dominate over nuclear stopping. TDTR results show that thermal transport in this sub-surface domain irradiated by 710 MeV ions with the dose of 1012 cm-2 has degraded by the factor of ~ 2 compared to non-irradiated sample.
Theoretical analysis was based on Klemens analytical thermal transport model [3] by fitting into TDTR measured thermal conductivity values. Given TDTR probed sub-surface region overlapped with ionization depth determined by SRIM calculation, we conclude that defects are formed primarily by electronic energy loss / ionization. Measurement and simulation of nano-scale thermal transport degradation in irradiated LiF will be discussed with results of structural defects proliferation as determined by OA and PL measurements. Funding by MES RK grant AP05130446, state-targeted program BR05236454 and NU FDCRG grant 110119FD4501 is acknowledged.
References
[1] Manika I. et al. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 196, no. 3–4 (November 1, 2002): 299–307.
[2] Schwartz K. et al. Physical Review B 58, no. 17 (November 1, 1998): 11232–40.
[3] Klemens P.G. Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms 1, no. 2–3 (February 1, 1984): 204–8.