Todd Allen

Profile Information
Publications:
"2.6 MeV proton irradiation effects on the surface integrity of depleted UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Andrew Nelson, Janne Pakarinen, Nuclear Instruments and Methods B Vol. 319 2014 100-106 Link
The effect of low temperature proton irradiation in depleted uranium dioxide was examined as a function of fluence. With 2.6 MeV protons, the fluence limit for preserving a good surface quality was found to be relatively low, about 1.4 and 7.0 × 1017 protons/cm2 for single and poly crystalline samples, respectively. Upon increasing the fluence above this threshold, severe surface flaking and disintegration of samples was observed. Based on scanning electron microscopy (SEM) and X-ray diffraction (XRD) observations the causes of surface failure were associated to high H atomic percent at the peak damage region due to low solubility of H in UO2. The resulting lattice stress is believed to exceed the fracture stress of the crystal at the observed fluencies. The oxygen point defects from the displacement damage may hinder the H diffusion and further increase the lattice stress, especially at the peak damage region.
"Advanced Test Reactor National Scientific User Facility: Addressing Advanced Nuclear Materials Research" Todd Allen, James Cole, John Jackson, Frances Marshall, INL/CON-12-27737 Vol. 2013 Link
The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.
"Bubble Character, Kr Distribution and Chemical Equilibrium in UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Hunter Henderson, Michele Manuel, Andrew Nelson, Janne Pakarinen, Billy Valderrama, Journal of Nuclear Materials Vol. 2015 Link
"Bubble evolution in Kr-irradiated UO2 during annealing" Lingfeng He, Xianming Bai, Janne Pakarinen, Brian Jaques, Jian Gan, Andrew Nelson, Anter EL-AZAB, Todd Allen, Journal of Nuclear Materials Vol. 496 2017 242-250 Link
Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO2 annealed at high temperature was conducted in order to understand the inert gas behavior in oxide nuclear fuel. The average diameter of intragranular bubbles increased gradually from 0.8 nm in as-irradiated sample at room temperature to 2.6 nm at 1600 °C and the bubble size distribution changed from a uniform distribution to a bimodal distribution above 1300 °C. The size of intergranular bubbles increased more rapidly than intragranular ones and bubble denuded zones near grain boundaries formed in all the annealed samples. It was found that high-angle grain boundaries held bigger bubbles than low-angle grain boundaries. Complementary atomistic modeling was conducted to interpret the effects of grain boundary character on the Kr segregation. The area density of strong segregation sites in the high-angle grain boundaries is much higher than that in the low angle grain boundaries.
"Bubble formation and Kr distribution in Kr-irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Andrew Nelson, Janne Pakarinen, Billy Valderrama, Lingfeng He, Abdel-Rahman Hassan, Hunter Henderson, Marquis Kirk, Michele Manuel, Journal of Nuclear Materials Vol. 456 2015 125-132 Link
In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO2 matrix and high release of Kr from sample surface under irradiation.
"Bubble, stoichiometry, and chemical equilibrium of krypton-irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Michele Manuel, Janne Pakarinen, Billy Valderrama, Abdel-Rahman Hassan, Marquis Kirk, Andrew Nelson, Journal of Nuclear Materials Vol. 456 2015 125-132 Link
In situ and ex situ transmission electron microscopy observation of small Kr bubbles in bothsingle-crystal and polycrystalline UO2 were conducted to understand the inert gas bubblebehavior in oxide nuclear fuel. The bubble size and volume swelling are shown as a weakfunction of ion dose but strongly depend on the temperature. The Kr bubble formation at roomtemperature was observed for the first time. The depth profiles of implanted Kr determined byatom probe tomography are in good agreement with the calculated profiles by SRIM, but themeasured concentration of Kr is about 1/3 of calculated one. This difference is mainly due to lowsolubility of Kr in UO2 matrix, which has been confirmed by both density-functional theorycalculations and chemical equilibrium analysis.
"Cesium and Silver Diffusion in SiC for TRISO Applications" Todd Allen, Tyler Gerczak, Zihua Zhu, Transactions of the American Nuclear Society Vol. 104 2011 47-48 Link
"Characterization of microstructure and property evolution in advanced cladding and duct: Materials exposed to high dose and elevated temperature" Todd Allen, Zhijie Jiao, Djamel Kaoumi, Janelle Wharry, cem topbasi, Aaron Kohnert, Leland Barnard, Alicia Certain, Kevin Field, Gary Was, Dane Morgan, Arthur Motta, Brian Wirth, Yong Yang, Journal of Materials Research Vol. 30 2015 1246-1274 Link
Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.
"Effect of exposure environment on surface decomposition of SiC–silver ion implantation diffusion couples" Todd Allen, Kevin Field, Tyler Gerczak, Guiqiu Zheng, Journal of Nuclear Materials Vol. 456 2015 281-286 Link
SiC is a promising material for nuclear applications and is a critical component in the construction of tristructural isotropic (TRISO) fuel. A primary issue with TRISO fuel operation is the observed release of 110mAg from intact fuel particles. The release of Ag has prompted research efforts to directly measure the transport mechanism of Ag in bulk SiC. Recent experimental efforts have focused primarily on Ag ion implantation designs. The effect of the thermal exposure system on the ion implantation surface has been investigated. Results indicate the utilization of a mated sample geometry and the establishment of a static thermal exposure environment is critical to maintaining an intact surface for diffusion analysis. The nature of the implantation surface and its potential role in Ag diffusion analysis are discussed.
"Effect of exposure environment on surface decomposition of SiC–silver ion implantation diffusion couples" Tyler Gerczak, Guiqiu Zheng, Kevin Field, Todd Allen, Journal of Nuclear Materials Vol. 456 2015 281-286 Link
"Effect of Grain Boundaries on Krypton Segregation Behavior in Irradiated Uranium Dioxide" Todd Allen, Darryl Butt, Jian Gan, Lingfeng He, Hunter Henderson, Brian Jaques, Michele Manuel, Janne Pakarinen, Billy Valderrama, Journal of Metals Vol. 66 2014 2562-2568 Link
Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregating toward grain boundaries and eventually leading to a lowering in thermal conductivity and fuel swelling. Recent computational studies have identi?ed that differences in grain boundary structure have a signi?cant effect on the segregation behavior of fission products. However, experimental work supporting these simulations is lacking. Atom probe tomography was used to measure the Kr distribution across grain boundaries in UO2. Polycrystalline depleted UO2 samples were irradiated with 0.7 MeV and 1.8 MeV Kr-ions and annealed to 1000C, 1300C, and 1600C for 1 h to produce a Kr-bubble dominated microstructure. The results of this work indicate a strong dependence of Kr concentration as a function of grain boundary structure. Temperature also influences grain boundary chemistry with greater Kr concentration evident at higher temperatures, resulting in a reduced Kr concentration in the bulk. Although Kr segregation takes place at elevated temperatures, no change in grain size or texture was observed in the irradiated UO2 samples.
"In Situ TEM Observation of Dislocation Evolution in Polycrystalline UO2" Todd Allen, Jian Gan, Mahima Gupta, Janne Pakarinen, Lingfeng He, Marquis Kirk, JOM Vol. 66 2014 2553-2561 Link
In situ transmission electron microscopy observation of polycrystalline UO2 (with average grain size of about 5 µm) irradiated with Kr ions at 600°C and 800°C was conducted to understand the radiation-induced dislocation evolution under the influence of grain boundaries. The dislocation evolution in the grain interior of polycrystalline UO2 was similar under Kr irradiation at different ion energies and temperatures. As expected, it was characterized by the nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation lines and tangles at high doses. For the first time, a dislocation-denuded zone was observed near a grain boundary in the 1-MeV Kr-irradiated UO2 sample at 800°C. The denuded zone in the vicinity of grain boundary was not found when the irradiation temperature was at 600°C. The suppression of dislocation loop formation near the boundary is likely due to the enhanced interstitial diffusion toward grain boundary at the high temperature.
"In-Situ TEM Observation of Dislocation Evolution in Kr-Irradiated UO2 Single Crystal" Todd Allen, Jian Gan, Mahima Gupta, Janne Pakarinen, Clarissa Yablinsky, Marquis Kirk, Xianming Bai, Journal of Nuclear Materials Vol. 443 2013 71-77 Link
In situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 were similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 °C and 1 MeV at 800 °C) used in this work. Interstitial-type dislocation loops with Burgers vector along 〈1 1 0〉 were observed in the Kr-irradiated UO2. The irradiated specimens were denuded of dislocation loops near the surface.
"Investigation of material property influenced stoichiometric deviations as evidenced during UV laser-assisted atom probe tomography in fluorite oxides" Todd Allen, Jian Gan, Hunter Henderson, Michele Manuel, Billy Valderrama, Clarissa Yablinsky, Nuclear Instruments and Methods in Physics Research B: Beam Interactions with Materials and Atoms Vol. 359 2015 107-114 Link
Oxide materials are used in numerous applications such as thermal barrier coatings, nuclear fuels, and electrical conductors and sensors, all applications where nanometer-scale stoichiometric changes can affect functional properties. Atom probe tomography can be used to characterize the precise chemical distribution of individual species and spatially quantify the oxygen to metal ratio at the nanometer scale. However, atom probe analysis of oxides can be accompanied by measurement artifacts caused by laser-material interactions. In this investigation, two technologically relevant oxide materials with the same crystal structure and an anion to cation ratio of 2.00, pure cerium oxide (CeO2) and uranium oxide (UO2) are studied. It was determined that electronic structure, optical properties, heat transfer properties, and oxide stability strongly affect their evaporation behavior, thus altering their measured stoichiometry, with thermal conductance and thermodynamic stability being strong factors.
"Irradiation response of delta ferrite in as-cast and thermally aged cast stainless steel" Todd Allen, Yiren Chen, Zhangbo Li, Wei-Yang Lo, Janne Pakarinen, Yaqiao Wu, Yong Yang, Journal of Nuclear Materials Vol. 466 2015 201-207 Link
To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ~315 °C to 0.08 dpa (5.6 × 1019 n/cm2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10-9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.
"Measurement of Helium Generation in AISI 304 Reflector and Blanket Assemblies after Long-Term Irradiation in EBR-II" Frank Garner, B. M. Oliver, L. R. Greenwood, Douglas Porter, Todd Allen, Journal of ASTM International,Research Vol. 20 2013 Link
"Microstructural characterization and density change of 304 stainless steel reflector blocks after long-term irradiation in EBR-II" Yina Huang, J.M.K. Wiezorek, Frank Garner, Paula Freyer, T. Okita, M. Sagisaka, Y. Isobe, Todd Allen, Journal of Nuclear Materials Vol. 465 2015 516-530 Link
While thin reactor structural components such as cladding and ducts do not experience significant gradients in dpa rate, gamma heating rate, temperature or stress, thick components can develop strong local variations in void swelling and irradiation creep in response to gradients in these variables. In this study we conducted microstructural investigations by transmission electron microscopy of two 52 mm thick 304-type stainless steel hex-blocks irradiated for 12 years in the EBR-II reactor with accumulated doses ranging from ∼0.4 to 33 dpa. Spatial variations in the populations of voids, precipitates, Frank loops and dislocation lines have been determined for 304 stainless steel sections exposed to different temperatures, different dpa levels and at different dpa rates, demonstrating the existence of spatial gradients in the resulting void swelling. The microstructural measurements compare very well with complementary density change measurements regarding void swelling gradients in the 304 stainless steel hex-block components. The TEM studies revealed that the original cold-worked-state microstructure of the unirradiated blocks was completely erased by irradiation, replaced by high densities of interstitial Frank loops, voids and carbide precipitates at both the lowest and highest doses. At large dose levels the amount of volumetric void swelling correlated directly with the gamma heating gradient-related temperature increase (e.g. for 28 dpa, ∼2% swelling at 418 °C and ∼2.9% swelling at 448 °C). Under approximately iso-thermal local conditions, volumetric void swelling was found to increase with dose level (e.g. ∼0.2% swelling at 0.4 dpa, ∼0.5% swelling at 4 dpa and ∼2% swelling at 28 dpa). Carbide precipitate formation levels were found to be relatively independent of both dpa level and temperature and induced a measurable densification. Void swelling was dominant at the higher dose levels and caused measurable decreases in density. Void swelling at the lowest doses was larger than might be expected based on the dpa level, an observation in agreement with earlier studies showing that the onset of void swelling is accelerated by decreasing dpa rates.
"Microstructure changes and thermal conductivity reduction in UO2 following 3.9 MeV He2+ ion irradiation" Anter EL-AZAB, Jian Gan, Marat Khafizov, Andrew Nelson, Janne Pakarinen, Chris Wetteland, Lingfeng He, David Hurley, Todd Allen, Journal of Nuclear Materials Vol. 454 2014 283-289 Link
The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in X-ray diffraction after ion irradiation up to 5 × 1016 He2+/cm2 at low-temperature (<200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 μm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 μm) in the sample subjected to 5 × 1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9 × 1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55% for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.
"Microstructure evolution in Xe-irradiated UO2 at room temperature" Todd Allen, Anter EL-AZAB, Jian Gan, Lingfeng He, Janne Pakarinen, Marquis Kirk, Andrew Nelson, Xianming Bai, Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms Vol. 330 2014 55-60 Link
In situ Transmission Electron Microscopy was conducted for single crystal UO2 to understand the microstructure evolution during 300 keV Xe irradiation at room temperature. The dislocation microstructure evolution was shown to occur as nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation segments and tangles at higher doses. Xe bubbles with dimensions of 1-2 nm were observed after room-temperature irradiation. Electron Energy Loss Spectroscopy indicated that UO2 remained stoichiometric under room temperature Xe irradiation.
"Near Surface Stoichiometry in UO2: A Density Functional Theory Study" Todd Allen, Hunter Henderson, Michele Manuel, Billy Valderrama, Journal of Chemistry Vol. 2015 2015 1-8 Link
The mechanisms of oxygen stoichiometry variation in UO2 at different temperature and oxygen partial pressure are important for understanding the dynamics of microstructure in these crystals. However, very limited experimental studies have been performed to understand the atomic structure of UO2 near surface and defect effects of near surface on stoichiometry in which the system can exchange atoms with the external reservoir. In this study, the near (110) surface relaxation and stoichiometry in UO2 have been studied with density functional theory (DFT) calculations. On the basis of the point-defect model (PDM), a general expression for the near surface stoichiometric variation is derived by using DFT total-energy calculations and atomistic thermodynamics, in an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2. By using the derived expression, it is observed that, under poor oxygen conditions, the stoichiometry of near surface is switched from hyperstoichiometric at 300K with a depth around 3 nm to near-stoichiometric at 1000K and hypostoichiometric at 2000 K. Furthermore, at very poor oxygen concentrations and high temperatures, our results also suggest that the bulk of the UO2 prefers to be hypostoichiometric, although the surface is near-stoichiometric.
"Observations of Ag diffusion in ion implanted SiC" Todd Allen, Tyler Gerczak, bin leng, Kumar Sridharan, Jerry Hunter, Andrew Giordani, Journal of Nuclear Materials Vol. 461 2015 314-324 Link
The nature and magnitude of Ag diffusion in SiC has been a topic of interest in connection with the performance of tristructural isotropic (TRISO) coated particle fuel for high temperature gas-cooled nuclear reactors. Ion implantation diffusion couples have been revisited to continue developing a more complete understanding of Ag fission product diffusion in SiC. Ion implantation diffusion couples fabricated from single crystal 4H-SiC and polycrystalline 3C-SiC substrates and exposed to 1500–1625 °C, were investigated by transmission electron microscopy and secondary ion mass spectrometry (SIMS). The high dynamic range of SIMS allowed for multiple diffusion régimes to be investigated, including enhanced diffusion by implantation-induced defects and grain boundary (GB) diffusion in undamaged SiC. Estimated diffusion coefficients suggest GB diffusion in bulk SiC does not properly describe the release observed from TRISO fuel.
"Radiation-resistant nanotwinned austenitic stainless steel" Gabriel Meric, I.M. Robertson, Todd Allen, Jean-Claude van Duysen, Kumar Sridharan, Scripta Materialia Vol. 159 2019 123-127 Link
A key strategy to increase the radiation resistance of materials has been to introduce a high density of interfaces that can act as sinks for radiation-induced defects. Twin boundaries are a type of interface that can be introduced through deformation but are usually considered to be ineffective sinks. Using heavy ion irradiation and transmission electron microscopy, this study investigates the influence of a high area per unit volume of twin boundaries on the radiation-induced swelling response of an austenitic stainless steel. The study shows that swelling can be suppressed in regions containing a high density of closely-spaced deformation twin boundaries.
"Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9wt.% Cr model ferritic/martensitic steel" Todd Allen, Heather Chichester, Kevin Field, Brandon Miller, Kumar Sridharan, Journal of Nuclear Materials Vol. 445 2013 143-148 Link
Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.
"Superior radiation-resistant nanoengineered austenitic 304L stainless steel for applications in extreme radiation environments" Cheng Sun, Lin Shao, Steven Zinkle, Todd Allen, Haiyan Wang, Xinghang Zhang, Scientific Reports Vol. 5 2015 Link
"Transmission Electron Microscopy Investigation of Krypton Bubbles in Polycrystalline CeO2" Todd Allen, Jian Gan, Mahima Gupta, Clarissa Yablinsky, Marquis Kirk, Nuclear Technology Vol. 182 2013 164-169 Link
To gain an understanding of gas bubble transport in oxide nuclear fuel, this paper uses polycrystalline CeO2, composed of both nanograins and micrograins, as a surrogate material for UO2. The CeO2 was implanted with 150-keV Kr ions up to a dose of 1 &times; 1016 ions/cm2 at 600&deg;C. Transmission electron microscopy characterizations of small Kr bubbles in nanograin and micrograin regions were compared. The grain boundary acted as an efficient defect sink, as evidenced by smaller bubbles and a lower bubble density in the nanograin region as compared to the micrograin region.
"Using a spherical crystallite model with vacancies to relate local atomic structure to irradiation defects in ZrC and ZrN" Todd Allen, Hasitha Ganegoda, Daniel Olive, Jeff Terry, Yong Yang, Clayton Dickerson, Journal of Nuclear Materials Vol. 475 2016 123-131 Link
Zirconium carbide and zirconium nitride are candidate materials for new fuel applications due to several favorable physicochemical properties. ZrC and ZrN samples were irradiated at the Advanced Test Reactor National Scientific User Facility with neutrons at 800 °C to a dose of 1 dpa. Structural examinations have been made of the ZrC samples using high resolution transmission electron microscopy, and the findings compared with a previous study of ZrC irradiated with protons at 800 °C. The use of X-ray absorption fine structure spectroscopy (XAFS) to characterize the radiation damage was also explored including a model based on spherical crystallites that can be used to relate EXAFS measurements to microscopy observations. A loss of coordination at more distant coordination shells was observed for both ZrC and ZrN, and a model using small spherical crystallites suggested this technique can be used to study dislocation densities in future studies of irradiated materials.
Presentations:
"Comparison of Computationally Simulated Fission Product Distribution with Correlative Characterization Techniques in Surrogate Nuclear Fuel Materials" Todd Allen, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Billy Valderrama, 2013 SACNAS National Conference October 2-6, (2013)
"Damage Structure Evolution in Ion Irradiated UO2" Todd Allen, Jian Gan, Mahima Gupta, Andrew Nelson, Jeff Terry, TMS 2014 February 16-20, (2014)
"Evolution of the ATR NSUF in Supporting Nuclear Fuels and Materials R&D" Todd Allen, James Cole, John Jackson, Frances Marshall, TMS 2014 February 16-20, (2014)
"Fission Products in Nuclear Fuel: Comparison of Simulated Distribution with Correlative Characterization Techniques" Todd Allen, Anter EL-AZAB, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Billy Valderrama, Clarissa Yablinsky, Microscopy and Microanalysis August 4-8, (2012)
"Heat treatment Effects on Precipitation in Irradiated HT9 Steel" Theresa Mary Green, Li He, Todd Allen, MiNES (Materials in Nuclear Energy Systems) 2019 October 6-10, (2019)
"Ion irradiation for nuclear materials research at University of Wisconsin-Madison" Li He, Gabriel Meric, Kim Kriewaldt, Kumar Sridharan, Adrien Couet, Todd Allen, The 51st Symposium of the North Eastern Accelerator Personnel September 23-27, (2018)
"Irradiation Effects in Oxide Nanoparticle Stability in Oxide Dispersion Strengthened (ODS) Steel" Todd Allen, Jianchao HE, Alexander Mairov, Kumar Sridharan, The Metallurgical Society Meeting March 15-19, (2015)
"Kr and Xe Bubble Characterization in CeO2" Todd Allen, Jian Gan, Mahima Gupta, Janne Pakarinen, TMS 2014 February 16-20, (2014)
"Microstructural Investigation of Kr Irradiated UO2" Todd Allen, Jian Gan, Mahima Gupta, Lingfeng He, Clarissa Yablinsky, The Minerals, Materials, and Metals Society, 2013 Annual Meeting & Exhibition March 3-7, (2013)
"Microstructural Investigations of Kr and Xe Irradiated UO2" Todd Allen, Anter EL-AZAB, Jian Gan, Mahima Gupta, Lingfeng He, Hunter Henderson, Michele Manuel, Janne Pakarinen, Billy Valderrama, Energy Frontier Research Centers Principal Investigators Meeting July 18-19, (2013)
"Microstructure Evolution in Dual Ion irradiated HT9 at 445 C and 460 C to 16.6 dpa" Li He, Theresa Mary Green, Todd Allen, MiNES (Materials in Nuclear Energy Systems) 2019 October 6-10, (2019)
"Nano-scale Irradiation Induced Chemistry Changes in Oxide" Todd Allen, Jian Gan, Lingfeng He, Hunter Henderson, Michele Manuel, Janne Pakarinen, Billy Valderrama, 2014 TMS Annual Meeting February 16-20, (2014)
"Nano-scale Irradiation Induced Chemistry Changes in Oxide Fuel Materials" Todd Allen, Jian Gan, Hunter Henderson, Janne Pakarinen, Billy Valderrama, TMS 2014 February 16-20, (2014)
"Poster - Examining microstructural differences in irradiated HT9 correlated with differences in processing prior to irradiation" Theresa Mary Green, Li He, Todd Allen, Brandon Miller, Lingfeng He, NUMAT 2018 October 15-18, (2018)
"Radiation Effects in UO2" Todd Allen, Jian Gan, Mahima Gupta, Michele Manuel, Andrew Nelson, Janne Pakarinen, Billy Valderrama, TMS 2014 February 16-20, (2014)
"Study of Interfacial Interactions Using Thin Film Surface Modification" Todd Allen, Alexander Mairov, Kumar Sridharan, MS&T Conference October 7-11, (2012)
"Void Swelling of Annealed 304 Stainless Steel at ~370-385&#61616;C and PWR-Relevant Displacement Rates" G.M> Bond, Bulent Sencer, Frank Garner, M.L. Hamilton, Todd Allen, Douglas Porter, 9th International Conference on Environmental Degradation of Materials in Nuclear Power Systems ? Water Reactors [unknown] Link
"Void Swelling of Annealed 304 Stainless Steel at ~370-385C and PWR-Relevant Displacement Rates" G.M. Bond, Francis Garner, Todd Allen, Douglas Porter, 9th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 1-2, (1999)
NSUF Articles:
U.S. DOE Nuclear Science User Facilities Awards 35 Rapid Turnaround Experiment Research Proposals - Awards total approximately $1.3 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE Office of Nuclear Energy. Wednesday, September 20, 2017 - Calls and Awards
RTE 1st Call Awards Announced - Projects total approximately $1.4 million These projects will continue to advance the understanding of irradiation effects in nuclear fuels and materials in support of the mission of the DOE-NE. Friday, February 8, 2019 - Calls and Awards
NSUF Researcher Feature: Kumar Sridharan - Learn more about a University of Wisconsin professor who helped kick start NSUF Sridharan's research team put the NSUF's first material samples into the ATR, launching a new era of research into the behaviors of fuels and materials in a nuclear reactor environment. Wednesday, August 28, 2019 - Newsletter, Researcher Highlight