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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Structure and Process of Managing the UO2 Dry In-Pile Fracture Test Irradiation Experiment

The management and structure of a novel research and development (R&D) project can be hard to navigate. This is especially true when it involves nuclear fuel. Familiarity with institutional safety and procedural requirements, a well defined scope, and accurate cost estimates can all help ensure successful project management from start to finish. The Dry In-Pile Fracture Test (DRIFT) experiments performed at the Transient Test Reactor Facility are an excellent example of managing an R&D project within its scope and budget constraints. The objective of the DRIFT test was to develop fracture propagation data in a manner consistent with light-water reactors to validate and improve the fracture propagation models of uranium oxide fuels in the MOOSE (Multiphysics Object-Oriented Simulation Environment)-BISON-Marmot code framework [1].

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Electrochemical Corrosion of SIMFUEL: Effects of dissolved H2 and noble metal particles

This report describes the results from electrochemical corrosion tests that were conducted using simulated spent fuel (SIMFUEL) materials comprised of UO2 and surrogate fission products. Two SIMFUEL compositions were tested to quantify the effect of noble metal inclusions and dissolved H2 on the UO2 dissolution rate. One material consisted of UO2 with added lanthanide oxides (UO2 N) and the other consisted of UO2 with added lanthanide oxides and noble metals (UO2-H) at concentrations to simulate high burnup fuel. The electrochemical corrosion tests on the SIMFUEL materials were conducted in aqueous electrolyte solutions (pH 10) that were either saturated with air or purged with a H2/Ar gas mixture to maintain a constant dissolved H2 concentration. The results from electrochemical corrosion tests on SIMFUEL can be applied to qualitatively understand and also quantify the separate effects of water chemistry and fuel composition on the degradation behavior of the UO2 matrix. Open circuit potential (OCP) measurements on a SIMFUEL material of known composition (i.e., known fraction of NMPs at the fuel surface) immersed in a known solution chemistry (i.e., pH, Eh, [O2], [H2]) at known temperature provide qualitative insight into the degradation behavior of the UO2 matrix. If the OCP is above the threshold potential at which the oxidative dissolution of U(IV) to U(VI) occurs, the SNF is expected to degrade by oxidative processes under the experimental conditions. The net currents that are measured during potentiostatic tests, during which the surface potential of the SIMFUEL material is fixed by a potentiostat, can be used to quantify the UO2 degradation rate and optimize the rate constant values used in the Fuel Matrix Degradation Model (FMDM) for half reactions that occur on the SNF surface. Specifically, electrochemical measurements enable the estimation of the total anodic current at ECORR–which is the surface potential at which the total anodic and total cathodic currents are equal–so that the rate constant values for key reactions can be calculated. The open circuit potential measurements, potentiostatic tests, surface property measurements (scanning electron microscopy images and electrochemical impedence spectroscopy plots), and solution elemental composition analyses are being performed to update and optimize the FMDM. A case study is presented herein to show how electrochemical corrosion test results and accompanying characterization results for the UO2 N material in air-saturated solution can be used to validate the SNF surface reaction module of the FMDM. Future electrochemical tests will be conducted to provide quantitative information on the effects of NM content (burnup), H2 concentration, water chemistry, temperature, and galvanic couples with cladding and EBS alloys on UO2 degradation that can be used to improve the accuracy and functionality of the FMDM.

Thomas, Sara↗
DOE OSTI2021

Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide

To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.

Uranium Dioxide, Gadolinium Oxide, Thermal Air Oxi↗
NASA NTRS1973

Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

Collins, J. F.↗
DOE OSTI2025

Hydrochlorination of Uranium Dioxide in a Molten Salt Mixture - Phase 1: Tube Furnace Cross Flow Experiments

In 2023, Metatomic® Inc., a South Carolina based company, was awarded a Gateway for Advanced Innovation in Nuclear (GAIN) research voucher for a proposed series of experiments aimed at demonstrating the viability of a spent nuclear fuel (SNF) recycling process patented by Met atomic® Inc. For the GAIN voucher, Metatomic® Inc. selected Savannah River National Laboratory (SRNL) as a partner in executing the proposed proof-of-concept experiments. This report outlines the proof-of-concept experiments performed by SRNL for Metatomic® Inc. during Phase 1 (of 2) experimentation. The Phase 1 hydrochlorination experiments consisted of weighing UO2 into alumina crucibles with a eutectic mixture of NaCl and CsCl, heating the uranium/salt mixture to a varied temperature (550, 650, or 750 °C), and flowing anhydrous hydrogen chloride (AHCl) gas across the surface of the uranium-bearing molten salt mixture. The hydrochlorination process conditions were maintained for 4-5 hours and the percent conversion for each batch of UO2 was determined using a suite of analytical characterization techniques. The degree of UO2 conversion was found to be greatest at the highest tested temperature, ultimately achieving 32.9% conversion of UO2 to water soluble uranium-chloro species (e.g., UO2Cl2, UCl4) after exposure to AHCl for 4.25 hours at 750 °C.

Nguyen, Vinh T. [Savannah River National Laborator↗
DOE OSTI2025

Ligand‐Directed Actinide Oxo‐Bond Manipulation in Actinyl Thiacalix[4]arene Complexes

Understanding the chemistry of the inert actinide oxo bond in actinyl ions AnO2 2+ is important for controlling actinide behavior in the environment, during separations, and in nuclear waste (An=U, Np, Pu). The thioether calixarene TC4A (4-tert-butyltetrathiacalix[4]arene) binds equatorially to the actinyl cation forming a conical pocket that differentiates the two trans-oxo groups. The 'ate' complexes, [A]2[UO2(TC4A)] (A=[Li(DME)2], HNEt3) and [HNEt3]2[AnO2(TC4A)] (An=U, Np, Pu), enable selective oxo chemistry. Silylation of the UVI oxo groups by bis(trimethylsilyl)pyrazine occurs first at only the unencapsulated exo oxo and only one silylation is needed to enable migration of the endo oxo out of the cone, whereupon a second silylation affords the stable UIV cis-bis(siloxide) [A]2[U(OSiMe3)2(TC4A)]. Calculations confirm that only one silylation event is needed to initiate oxo rearrangement, and that the putative cis dioxo isomer of [UO2(TC4A)]2- would be stable if it could be accessed synthetically, at only 23 kcal.mol-1 in energy above the classical trans dioxo. Calculations for the transuranic cis[AnO2(TC4A)]2- (An=Np, Pu) are at higher energies, 30-35 kcal.mol-1, retaining the U complexes as the more obvious target for a cis-dioxo actinyl ion. The aryloxide (OAr) groups of the macrocycle are essential in stabilizing this as-yet unseen uranyl geometry as further bonding in the TC4A U-OAr groups stabilizes the U=O 'yl' bonds, explaining the stability of the putative cis[UO2(TC4A)]2- in this ligand framework.

Pyrch, Mikaela M↗
DOE OSTI2023

In-Situ Transmission Electron Microscopy Study of the Evolution of Extended Defects in Oxide Nuclear Fuels

In-situ transmission electron microscopy (TEM) with simulated extreme environments is an effective tool for understanding and tracking microstructural changes down to the atomic scale. The objective of this research is to study the effect of temperature on the evolution of extended defects in ThO2 and UO2. Here, we present in-situ TEM isochronal thermal annealing experiments using a micro-electro-mechanical-system (MEMS)-based heating holder. ThO2 and UO2 single crystal specimens were grown inside an inert silver ampoule using hydrothermal synthesis. Both samples were irradiated using 2 MeV protons at 600oC up to 0.1 dpa at Texas A&M University’s Accelerator Laboratory. Fig. 1a shows a weak-beam dark field TEM image of ThO2 after irradiation, indicating the presence of faulted 1/3 <111> type dislocation loops. This presentation will discuss the effect of annealing temperature (600oC, 800oC, 1000oC, 1100oC, etc.) on dislocation loop density, size and distribution, loop nature (interstitial/vacancies) and Burgers vector, as well as the formation of voids in both ThO2 and UO2. This work will also discuss the interaction between defects during annealing. The in-situ TEM annealing cycle is shown in Fig 1b. This research significantly improves the understanding of defect behavior in oxide nuclear fuels with temperature and will aid computational modeling efforts. This work was supported as part of the Center for Thermal Energy Transport under Irradiation (TETI) Energy Frontier Research Center, funded by the U.S. Department of Energy Office of Science.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on U(HO2)2 by Materials Project

UO2(OH)2 crystallizes in the triclinic P1 space group. The structure is two-dimensional and consists of one molecular hydrogen;dihydrate molecule and one UO2 sheet oriented in the (0, 0, 1) direction. In the UO2 sheet, U6+ is bonded in a square co-planar geometry to four O2- atoms. There are a spread of U–O bond distances ranging from 2.05–2.11 Å. There are two inequivalent O2- sites. In the first O2- site, O2- is bonded in a linear geometry to two equivalent U6+ atoms. In the second O2- site, O2- is bonded in a linear geometry to two equivalent U6+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Destructive PIE and Safety Testing of Six AGR-2 UO 2 Capsule 3 Compacts

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2) was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from June 2010 to October 2013 (Collin 2014). The fuel compacts in this experiment held either tristructural isotropic (TRISO)-coated spherical kernels of uranium oxide (UO2) or TRISO-coated kernels containing both uranium carbide and uranium oxide phases (UCO). There were six separately monitored and controlled capsules in the AGR-2 test train. Capsule 3 held twelve compacts containing UO2-TRISO particles fabricated by BWX Technologies Nuclear Operations Group. The AGR-2 TRISO particles were fabricated in a pilot-scale fluidized-bed chemical vapor deposition (FB-CVD) furnace with a coating chamber inner diameter of 150 mm (Phillips, Barnes, and Hunn 2010), which was a change from the first irradiation experiment (AGR-1) particles that had been coated in a lab-scale FB-CVD coating system with a chamber inner diameter of 50 mm (Lowden 2006). The TRISO particles were overcoated with resinated graphite flake at Oak Ridge National Laboratory (ORNL), and the overcoated particles were pressed into one-inch-long, half-inch-diameter cylinders (Hunn, Montgomery, and Pappano 2010). Each cylindrical compact held an average of 1,543 TRISO particles with 9.6% enriched UO2 kernels that had a nominal diameter of 500 μm (Hunn, Savage, and Silva 2012). Capsule 3 compacts were irradiated to average calculated burnups of 9.01–10.69% fissions per initial metal atom (FIMA), and the average calculated fluences of fast neutrons with energies E n > 0.18 MeV were 3.05–3.53×10 25 n/m 2 (Sterbentz 2014). The calculated time-average, volume-average Capsule 3 compact temperatures were 996–1,062°C. However, Capsule 3 compact temperatures varied several hundred degrees across each compact, and the timeaverage minimum (TA min ) and time-average maximum (TA max ) temperatures were between 889–999°C and 1,072–1,105°C, respectively (Hawkes 2014). After irradiation, the AGR-2 test train was transferred from ATR to the INL Materials and Fuels Complex for inspection and disassembly (Ploger, Demkowicz, and Harp 2015). The initial inspection included dimensional metrology of the compacts and graphite fuel holders. Like all the AGR-2 compacts, the compacts in Capsule 3 shrank slightly during irradiation, as expected, with an average length reduction of 1.07–1.24% and an average diameter reduction of 0.13–0.41%. Post-irradiation examination (PIE) of the capsule components was completed to measure select fission products ( 90 Sr, 110m Ag, 134 Cs, 137 Cs, 144 Ce, and 154 Eu) outside the compacts (Stempien and Demkowicz 2020). This involved gamma counting of the graphite and graphoil spacers at the top and bottom of each capsule, acid leaching for radiochemical analysis of fission products on the metallic capsule components, and burn-leach analysis of the graphite holders. The total amount of 110mAg measured on the Capsule 3 components was 13% of the calculated capsule inventory. This was significantly lower than the amount of 110m Ag measured on the three UCO capsule components, which ranged from 32–70%. The lower 110m Ag release in Capsule 3 was likely due to lower peak temperatures compared with the UCO fuel capsules (Hawkes 2014). Measured inventories of the other select fission products were also lower in Capsule 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Evaluation of refractory-metal-clad uranium nitride and uranium dioxide fuel pins after irradiation for times up to 10 450 hours at 990 C

The effects of some materials variables on the irradiation performance of fuel pins for a lithium-cooled space power reactor design concept were examined. The variables studied were UN fuel density, fuel composition, and cladding alloy. All pins were irradiated at about 990 C in a thermal neutron environment to the design fuel burnup. An 85-percent dense UN fuel gave the best overall results in meeting the operational goals. The T-111 cladding on all specimens was embrittled, possibly by hydrogen in the case of the UN fuel and by uranium and oxygen in the case of the UO2 fuel. Tests with Cb-1Zr cladding indicate potential use of this cladding material. The UO2 fueled specimens met the operational goals of less than 1 percent cladding strain, but other factors make UO2 less attractive than low-density UN for the contemplated space power reactor use.

Bowles, K. J.↗
NASA NTRS2014

Design Evolutuion of Hot Isotatic Press Cans for NTP Cermet Fuel Fabrication

Nuclear Thermal Propulsion (NTP) is under consideration for potential use in deep space exploration missions due to desirable performance properties such as a high specific impulse (> 850 seconds). Tungsten (W)-60vol%UO2 cermet fuel elements are under development, with efforts emphasizing fabrication, performance testing and process optimization to meet NTP service life requirements [1]. Fuel elements incorporate design features that provide redundant protection from crack initiation, crack propagation potentially resulting in hot hydrogen (H2) reduction of UO2 kernels. Fuel erosion and fission product retention barriers include W coated UO2 fuel kernels, W clad internal flow channels and fuel element external W clad resulting in a fully encapsulated fuel element design as shown.

Mireles, O. R.↗
NASA NTRS2015

A Historical Review of Cermet Fuel Development and the Engine Performance Implications

This paper reviews test data for cermet fuel samples developed in the 1960's to better quantify Nuclear Thermal Propulsion (NTP) cermet engine performance, and to better understand contemporary fuel testing results. Over 200 cermet (W-UO2) samples were tested by thermally cycling to 2500 deg (2770 K) in hydrogen. The data indicates two issues at high temperatures: the vaporization rate of UO2 and the chemical stability of UO2. The data show that cladding and chemical stabilizers each result in large, order of magnitude improvements in high temperature performance, while other approaches yield smaller, incremental improvements. Data is very limited above 2770 K, and this complicates predictions of engine performance at high Isp. The paper considers how this material performance data translates into engine performance. In particular, the location of maximum temperature within the fuel element and the effect of heat deposition rate are examined.

Ceramic Nuclear Fuel↗
DOE OSTI2021

Microstructural and Micro-Chemical Evolutions in the Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 TRISO fuel particles were fabricated with slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. To correlate those differences with the fuel kernel responses to neutron irradiations in terms of irradiated fuel microstructure, fission products chemical and physical states, and fission gas bubble evolutions, extensive microstructural and analytical characterizations were conducted. The studies used a state of art transmission electron microscopy (TEM) equipped with Energy-dispersive X-ray spectroscopy (EDS) of four silicon solid-state detectors which have super sensitivity and fast speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen to represent a representative irradiation conditions with a fuel burnup within the range from 10.8 to 18.6% FIMA, and the time-average volume-average temperatures vary from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% for the AGR1 fuel kernels and 14.03 wt.% for the AGR-2 fuel kernels. The TEM results show that there were significant microstructural reconstructions in the irradiated fuel kernels for both the AGR-1 and AGR-2 fuels. There are four major phases including fuel matrix of UO2 and UC, U2RuC2, and UMoC2 in the irradiated AGR2 fuel kernel. Zr and Nb form solid solution in the UC phase. UMoC2 phase often shows a detectable concentration of Tc. Pd was found to mainly locate in the buffer layer or to be associated with fission gas bubble within the UMoC2 phase. The EDS maps qualitatively show that the rare-earth fission products (Nb, et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there is a high number of rod-shape precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The difference of microstructural and micro-chemical evolutions in irradiated fuel kernels between the AGR-1 and AGR-2 TRISO fuel particle may result from a combined factor of irradiation temperature, fuel geometry and chemical composition. However, the irradiation temperature probably play a more deterministic role. Limited electron energy loss spectroscopy (EELS) characterizations on the AGR2 fuel kernel show that there is nearly no carbon in the UO2 phase while a small fraction of oxygen was detected in the UC/UMoC2 phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Design and validation of a gamma-ray scanning system for measuring irradiated nuclear fuel

We report a prototype gamma-ray scanning system was designed to perform post-irradiation measurements of nuclear fuel at Idaho National Laboratory (INL). The system is composed of a coaxial high-purity germanium (HPGe) detector, a collimator, and mechanical positioning stages that translate the fuel sample across the front of the collimator. This paper documents the design process, modeling and the laboratory characterization performed to validate the system. The modeling of the system was done using the Monte Carlo N-Particle code with vendor-supplied design specifications, dimensional measurements, and x-ray radiographs of the HPGe detector. The characterization included measurements of calibrated radiation sources in different geometries, and calculations for absolute efficiency and dead time. Benchmark measurements were performed scanning an irradiated fuel rodlet containing ten pellets of 4.9% enriched UO2 in zirconium alloy. These results were compared to simulated spectra to further help characterize the detector model. One goal of the system is to determine the number of fissions per gram of UO2 in the fuel; comparison of the results from this prototype system's assay of the irradiated fuel rodlet are in agreement with previously estimated results.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Application of HRGS for forensic characterization of uranium oxides, pure uranium metals and uranium alloys

A nondestructive iterative method for uranium-bearing material characterization with HRGS developed earlier in (J. appl. Rad. Isotopes, 166 (2020) 109433) is applied to determine matrix densities and uranium isotope masses of a variety of uranium materials, namely uranium ore, UO 2 and U 3 O 8 powders, fuel elements in the form of UO2 microspheres, uranium metal and uranium alloys. It is shown that U 3 O 8 powders with uranium mass fraction of about 84% can be distinguished from the powders of UO 2 with uranium mass fraction of about 87%; uranium products in the form of liquid or loose powder with matrix density of 0.5-2.0 g/cm 3 can be distinguished from uranium products in the form of compacted fuel elements with matrix density of 6.0-10.0 g/cm 3 and from pure metal uranium and uranium alloys with matrix density of 14.0- 19.0 g/cm 3 . In fuel microspheres based on UO2 the uranium mass fraction 88.02% measured by HRGS is consistent, within the measurement uncertainties, with the results of isotope dilution mass spectrometry 87.76±0.64 % and also is confirmed by X-ray diffraction technique. The uranium mass fraction of the uranium ore estimated as 0.08% by HRGS is consistent, within the measurement uncertainties, with the value 0.09±0.01% determined with WDXRF. Densities of two different uranium metal samples, estimated as 18.42 g/cm 3 and 19.33 g/cm 3 by HRGS are consistent with values 18.24±0.55 g/cm 3 and 18.86±0.59 g/cm 3 , respectively, obtained by the gas pycnometry technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Epsilon Metal: A Waste Form for Noble Metals from Used Nuclear Fuel

Epsilon metal (e-metal) is the metallic phase that forms as inclusions at the grain boundaries in the UO2 fuel during reactor operation. This metal is composed of Pd, Mo, Rh, Ru, and Tc. These metallic inclusions are insoluble in strong acid and remnants of these metallic inclusions have been found in the UO2 matrix that remains from the natural reactors in Gabon that were active 1.8 billion years ago, therefore e-metal should be an excellent waste form for the immobilization of the long-lived isotopes 107Pd (6.5 ? 106 a) and 99Tc (2.13 ? 105 a), with 99Tc being the isotope of interest for repository performance. Therefore, the chemical durability of this potential waste form is assessed in this study. Typically, corrosion rates for metallic materials are measured electrochemically because they are quick, inexpensive, and can reveal the mechanism by which a metal corrodes, at least initially. However, in a repository the waste form would be subjected to slowly flowing water without an applied electrical potential over long time periods. Therefore, the corrosion rates of e-metal specimens were measured with both electrochemical tests and the single-pass flow-through test (SPFT). Potentiodynamic and potentiostatic polarization results suggest that a thin passive film exists on the alloy surface, which seems to be responsible for its high corrosion resistance. Additionally, X-Ray photoelectron spectroscopic results suggest that Pd oxides are significantly enriched in the passive film Results from the SPFT show that the dissolution rates were weakly dependent on pH. Only Mo and Re were found in solution and were used for the calculation of the dissolution rates. In general, the electrochemically determined corrosion rates agree reasonably well with the initial dissolution rate measured with the SPFT test, but they are about one or two orders of magnitude higher than the steady state rates. The causes for this discrepancy are discussed.

epsilon-metal, nuclear waste form, corrosion, sing↗
DOE OSTI2021

Dislocation Loops in Proton Irradiated Uranium-Nitrogen-Oxygen System

Here in this study, we investigated the dislocation loop types formed in the proton-irradiated uranium-nitrogen-oxygen (U-N-O) system, which involves uranium mononitride (UN), uranium sesquinitride (a-U2N3), and uranium dioxide (UO2) phases. The dislocation loop formation is examined using specimens irradiated at 400°C and 710°C. Based on the detailed transmission-based electron microscopy characterization with i) the morphology-based on-zone and ii) the invisibility-criterion based two-beam condition imaging techniques, only a single type of dislocation loop in each phase is found: a/2?110?, a/2?111?, or a/3?111? dislocation loops in UN, a-U2N3, and UO2 phases, respectively. Molecular statics calculations for the formation energy of perfect and faulted dislocation loops in UN phases indicate a critical loop size of ~ 6 nm, above which perfect loops are energetically favorable. This could explain the absence of faulted loops in the experimental observation of the irradiated UN phase at two temperatures. This work will enhance the understanding of irradiation induced microstructural evolution for uranium nitrides as advanced nuclear fuels for the next-generation nuclear reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗