MICROSTRUCTURE EVOLUTION OF ZIRCONIUM CARBIDE IRRADIATED BY IONS

Transcription

1 The Pennsylvania State University The Graduate School Department of Mechanical and Nuclear Engineering MICROSTRUCTURE EVOLUTION OF ZIRCONIUM CARBIDE IRRADIATED BY IONS A Dissertation in Nuclear Engineering by Christopher Ulmer c 2014 Christopher Ulmer Submitted in Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy December 2014

2 The dissertation of Christopher Ulmer was read and approved 1 by the following: Arthur Motta Chair and Professor of Nuclear Engineering and Professor of Materials Science and Engineering Dissertation Advisor Chair of Committee Nasim Alem Assistant Professor of Materials Science and Engineering Kostadin Ivanov Distinguished Professor of Nuclear Engineering Igor Jovanovic Bashore Faculty Development Professorship Associate Professor of Nuclear Engineering Izabela Szlufarska Professor of Materials Science and Engineering and Engineering Physics at University of Wisconsin-Madison Special Member Karen Thole Department Head of Mechanical and Nuclear Engineering Professor of Mechanical Engineering 1 Signatures on file in the Graduate School.

3 iii Abstract ZrC is a candidate material for use in Generation IV high-temperature, gas-cooled reactor TRISO coated fuel particles, so it is important to understand its behavior under irradiation. The microstructural evolution of ZrC x under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Experiments were performed in which the sample stoichiometry and irradiation temperature were systematically varied. In situ experiments made it possible to continuously follow the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were methodically recorded at chosen dose points. Experiments centered on the irradiation of ZrC 0.8 and ZrC 0.9 with 1 MeV Kr ions at temperatures ranging from K up to 10 dpa. Initial damage developed as 2-4 nm diameter black-dot defects after a threshold dose of approximately dpa. As the irradiation temperature increased, the threshold dose for visible defect formation decreased. The density and size of defects increased with additional dose and the density of defects ranged on the order of m 3 for all experiments. The defect diameter also increased with irradiation temperature, with average defect diameters at 3 dpa ranging from approximately 4 nm at 673 K to 8 nm at 1073 K. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but agglomeration of small defects into loops occurred at 1073 K and resulted in an overall coarsening of the microstructure. The irradiated microstructure was found to not be strongly dependent

4 iv on the stoichiometry as results for the two stoichiometries studied were nearly identical. No irradiation induced amorphization was observed, even after 5 dpa at 20 K and 10 dpa at 50 K. At the higher temperature (873 K and above), the irradiated microstructure varied with sample thickness and showed a defect-denuded zone in the thin area near the edge. A one-dimensional cluster dynamics rate theory model that only considered the creation and mobility of point defects and their agglomeration into defect clusters was solved and compared with the experimental results. General trends from the simulation results matched the experimental observations: a threshold dose was predicted by the calculation, loop diameter was predicted to increases with dose and temperature, and loop density increased with dose and decreased with temperature, as observed. The spatial distribution showed lower loop size and density near the surface. Additional work is needed to match the experimental results quantitatively for both loop size and density, and the results were found to be sensitive to the chosen temperature.

5 v Table of Contents List of Tables vii List of Figures viii Acknowledgments xi Chapter 1. Introduction Motivation Zirconium Carbide ZrC for TRISO Radiation Damage in ZrC Goals of the Study Chapter 2. Experiment Material Sample Preparation Initial Characterization IVEM Irradiation Details Experimental Analysis Irradiation Dose Chapter 3. Characterization of Irradiated Microstructure Dose Dependence of Damage Accumulation Microstructure Evolution at 50 K and Below Microstructure Evolution at 300 K Microstructure Evolution at 873 K Microstructure Evolution at 1073 K Defect Measurement and Counting Temperature Dependence of Defect Morphology Habit Planes of Dislocation Loops at 1073K Dependence of Irradiated Microstructure on Sample Thickness Indexing Diffraction Rings/Spots Formed During Irradiation Effect of Stoichiometry on Damage Accumulation Thermal Annealing of ZrC 0.9 Irradiated to 1 DPA at 300 K Discussion Chapter 4. Modeling of Irradiation Induced Defect Evolution Rate Theory Model Method Results Interpretation

6 4.5 Discussion Chapter 5. Summary and Conclusions Microstructure Evolution Rate Theory Modeling Future Work Bibliography vi

7 vii List of Tables 1.1 Elastic constants of ZrC.[12] Formation energy [ev/defect] of point defects in ZrC [16] Measured stoichiometry for samples of ZrC x List of experiments conducted at the IVEM in this work Calculated extinction distances for ZrC Input parameters for SRIM simulation Defect formation [16] and migration energies [40] Parameters used in solution of the rate theory equations

8 viii List of Figures 1.1 TRISO fuel particle [4] The crystal structure of ZrC [16] Phase diagram of the Zr-C system [10] Interstitial defect configurations reported by Kim, et al. [16] TEM micrographs of the irradiated microstructure of ZrC observed at 0.7 and 1.5 dpa by Yang, et al. [31] TEM micrographs of the neutron irradiated microstructure of ZrC observed by Snead, et al. [18] at various temperatures Photograph of as-received ZrC rod Optical microscope images showing the general microstructure of ZrC x Optical microscope image showing the grains of ZrC Optical microscope images showing the two-phase structure of ZrC A 3g dark-field TEM micrograph showing the unniradiated microstructure of ZrC 0.9. Pre-existing dislocations are marked by small arrows A photograph of the IVEM (left) and a sketch of the ion beam and TEM interface[34] (right) Example of local thickness in ZrC 0.9 determined by thickness fringes in a g = 220 two-beam bright-field condition CBED pattern recorded with g = 220 in ZrC 0.9. The direct beam is on the left and the diffracted beam is on the right CBED analysis performed at three thicknesses to determine the extinction distance for g = 220 in ZrC SRIM calculated damage profile for 1 MeV Kr ions into 100 nm thick ZrC Integral pka spectrum as calculated by SRIM for ZrC 0.9 and 1 MeV Kr ions A series of bright-field micrographs showing ZrC 0.9 irradiated to 1, 3 and 5 dpa Dark-field micrograph of ZrC 0.9 irradiated to 2 dpa at 50 K centered on a pre-existing dislocation (marked by arrow). No defect-denuded zone is seen Microstructure of ZrC 0.9 irradiated to 10 dpa at 50 K. (a): bright-field micrograph. (b): dark-field micrograph using diffracted intensity from ZrC matrix. (c): dark-field micrograph using diffracted intensity from the brightest ring down and to the left from the beam stop. (d): selected area diffraction pattern Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 300 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa... 41

9 3.5 Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 300 K showing the microstructure at 3, 4, and 5 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 300 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 300 K showing the microstructure at 3, 4, and 5 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 873 K showing the microstructure at 0.05, 0.1, 0.3, and 0.5 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 873 K showing the microstructure at 1, 2, 3, and 4 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 873 K showing the microstructure at 0, 0.3, 0.5, and 1 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 873 K showing the microstructure at 2, 3, 4, and 5 dpa A diffraction pattern taken from ZrC 0.9 irradiated to 5 dpa at 873 K Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 1073 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 1073 K showing the microstructure at 3, 4, 6, 8, and 10 dpa Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 1073 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 1073 K showing the microstructure at 3, 4, and 5 dpa Series of dark-field TEM images captured from the video taken during the irradiation of ZrC 0.9 at 1073 K between 1 and 2 dpa showing the development of a dislocation loop. The images are centered on the dislocation loop location, and sketches of the loop shape, when discernible, are inset to the right Defect size distribution for ZrC 0.9 irradiated to 1 dpa at 673 K Quantitative measurements of the average defect diameter and defect density for ZrC 0.9 irradiated at 673 K, 873 K, and 1073 K Dark-field micrographs of ZrC 0.9 irradiated to 1 dpa at temperatures ranging from 300 K to 1073 K Dark-field micrographs of ZrC 0.9 irradiated to 3 dpa at temperatures ranging from 300 K to 1073 K Bright-field micrographs of ZrC 0.9 irradiated to 5 dpa (except where noted) at temperatures ranging from 300 K to 1073 K Diffraction pattern showing the zone axis near the imaging condition for Figure Dark-field micrograph of ZrC 0.9 irradiated to 2 dpa at 1073 K. Possible dislocation loop habit planes are indicated A bright-field micrograph of ZrC 0.9 irradiated to 4 dpa at 1073 K taken near the sample edge showing the variation in microstructure with sample thickness ix

10 3.26 Top: A bright-field micrograph of ZrC 0.8 irradiated to 5 dpa at 873 K taken near the sample edge. The marked region shows the analyzed area. Bottom: Quantitative analysis of the average (left) and maximum (right) defect diameter as a function of an arbitrary distance from the edge of the sample A cross-section schematic showing the proposed defect profile due to the action of the sample surface as a point defect sink Left: A diffraction pattern shows the diffraction rings that develop in ZrC 0.9 irradiated to 5 dpa at 873 K. Right: The diffracted intensity is integrated over a wedge from the DP and the expected FCC lattice peaks with a 0 = 5.06 Å are indicated Dark-field micrographs of ZrC 0.8 (left) and ZrC 0.9 (right) irradiated to 5 dpa at 473 K (top) and 1073 K (bottom) showing similar microstructures Dark-field TEM images showing the progression of the microstructure through the annealing steps at 473 K, 673K, 873 K and 1073 K Dark-field TEM images comparing the microstructures obtained by annealing at 1073 K after irradiation to 1 dpa at 300 K (top) to that obtained by irradiating to 1 dpa at 1073 K (bottom) Cluster dynamics simulation results showing the dose evolution of average loop size and density for irradiations at 300 K and 673 K Comparison of the defect density and average defect diameter results from cluster dynamics simulation and experiment Cluster dynamics simulation results showing the defect size distribution when the average loop diameter was 4 nm during irradiations at 300 K and 673 K Cluster dynamics simulation results showing the through-thickness profile of the vacancy concentration, interstitial concentration, average loop diameter and loop density at 673 K and dpa Evolution of interstitial concentration as a function of dose and temperature Cluster dynamics simulation results showing the dose evolution of average loop size and density for irradiations at 673 K and 1073 K Cluster dynamics simulation results showing the dose evolution of average loop size and density for irradiations at 200 K and 300 K x

11 xi Acknowledgments I would first like to thank my dissertation adviser, Arthur Motta, for his continuous help throughout the project, and his ability to always ask the difficult questions during our discussions. I would also like to thank the project collaborators at the University of Wisconsin- Madison, Izabela Szlufarska, Dane Morgan, Ming-Jie Zheng and Yina Huang, for their open dialogue to discuss research results. In addition, I would also like to thank my committee members at Penn State University, Nasim Alem, Kostadin Ivanov and Igor Jovanovic for their help in reviewing my dissertation. I also thank Mark Kirk, Pete Baldo and Edward Ryan from the IVEM at Argonne National Laboratory; they were instrumental in carrying out my experiments and obtaining good data. I thank all the members of the Nuclear Materials group at Penn State University for providing a great working environment and enlightening discussions. In particular, I thank Cem Topbasi for his help in all aspects of this research. Last but not least, I would like to thank my parents for their unwavering support.

12 Get your facts first, and then you can distort them as much as you please. Mark Twain xii

13 1 Chapter 1 Introduction 1.1 Motivation Generation IV nuclear reactors are a proposed set of revolutionary designs in reactor systems which seek improvements in sustainability, safety, reliability, economics, and proliferation resistance over current designs [1, 2]. Among the proposed reactor designs is a type of high-temperature gas-cooled reactor, the very-high-temperature reactor (VHTR). The reactor is designed to operate in the thermal neutron spectrum and to use a helium gas coolant with graphite moderator. The coolant outlet temperature is within the range of C in the near future and preferably 1000 C or higher if possible, since the high outlet temperature allows for both higher efficiency in electricity generation and for supplying process heat for chemical reactions such as required for hydrogen production. The fuel form of the VHTR is the coated fuel particle, in particular the tristructural isotropic (TRISO) fuel [3]. An example of a TRISO fuel particle is shown in Figure 1.1. The basic design includes a center fuel kernel of either UO 2 or UCO. Directly outside this layer is a porous, low-density buffer layer of carbon whose main functions include attenuating fission product recoils, providing expansion volume for fission gases, and acting as a sacrificial layer to accommodate kernel swelling. Around the buffer layer is the inner pyrolytic carbon (IPyC) layer; this layer provides a base for silicon carbide

14 2 (SiC) deposition, and during operation it can retain fission gases and help keep the SiC layer in compression. The next layer is the SiC layer which provides structural support for the fuel particle and is the primary barrier against fission product release. The final outer pyrolytic carbon (OPyC) layer compresses the SiC layer and provides a final barrier against the release of gaseous fission products. These coated fuel particles are finally suspended in a much larger graphite matrix which is either spherical or cylindrical in shape depending upon the specific reactor design. Figure 1.1. TRISO fuel particle [4]. There are some limitations for the use of SiC in the TRISO fuel design. For example, SiC is thermally unstable in the temperature range that may be reached during accident conditions [5]. In such an event, there could be a transition from the deposited β-sic to α-sic, resulting in a failure of the layer and the release of fission products. In addition, the SiC layer is susceptible to attack by the fission product palladium [6].

15 3 Palladium produced by fission in the fuel can diffuse through the fuel particle and corrode the inner surface of the SiC layer. With sufficient degradation of the SiC, the retention of fission products could be compromised. Because of these limitations, a variation on the standard TRISO fuel design was developed in which a zirconium carbide (ZrC) layer would be used in place of the SiC layer [7]. Because of its intended use as the load-bearing layer in TRISO fuel for a VHTR, the response of ZrC to irradiation damage must be understood. Energetic neutrons produced in a nuclear reactor environment can collide with atoms in the material producing localized atomic displacement cascades which result in the production of defect clusters and point defects such as vacancies and interstitial atoms. Further, these vacancies and interstitials can diffuse through the crystal lattice and interact with each other and the pre-existing microstructure to produce more complicated damage structures such as dislocation loops and networks. Local chemical compositions can be changed through irradiation induced segregation and precipitates can be formed or dissolved. These microscopic changes to the material may affect its macroscopic properties by such processes as void swelling, embrittlement, growth, or hardening. Because the macroscopic property changes are induced by the microscopic changes, an understanding of microstructural development during irradiation is key to understanding materials performance in a reactor environment. 1.2 Zirconium Carbide The Zr-C system has been studied both experimentally[8] and theoretically[9, 10]. ZrC exists over a large range of compositions, with the substoichiometric compound

16 4 remaining stable down to approximately ZrC 0.5. When the fraction of carbon exceeds 49.5% (ZrC 0.98 ), the extra carbon forms a second phase of free graphite that exists in equilibrium with the ZrC. ZrC melts congruently at 3700 K and 45.8% C and has a eutectic with graphite at 3200 K and 67.6% C. A phase diagram for the Zr-C system is shown in Figure 1.3. The ZrC phase takes on a cubic B1 crystal structure (similar to NaCl) as shown in Figure 1.2. This can be thought of as a face-centered cubic (FCC) Zr lattice where the carbon atoms take up positions at the octahedral interstitial sites. The metalmetal bonds of the base Zr metal are replaced with strong covalent σ-bonds and weaker metallic π-bonds between the Zr and C atoms [11]. This change in bonding results in a much greater resistance to shear in ZrC than seen in the base Zr metal. The elastic constants of ZrC have been investigated in several experiments, and values are in relatively good agreement [12, 13, 14]. A summary of these constants is given in Table 1.1. The measurements of micro-indentation hardness at room temperature show greater variation and ranged from GPa [15]. Table 1.1. Elastic constants of ZrC.[12] Young s Shear Bulk Poisson s Debye C 11 C 12 C 44 modulus modulus modulus ratio temperature [GPa] [GPa] [GPa] [GPa] [GPa] [GPa] [K] The properties of ZrC x have been shown to be sensitive to the stoichiometry of the material. For example, the thermal conductivity at 300 K was measured to be 6.6 W/mK at x=0.64 and increased to a maximum of 45.8 W/mK at x=0.982 [17]. In

17 5 Figure 1.2. The crystal structure of ZrC [16]. Figure 1.3. Phase diagram of the Zr-C system [10].

18 6 addition, the lattice parameter was shown to vary from Å at 38.5% C up to a maximum of Å at 45.5% C before decreasing again to Å at 49% C [8]. Additional examples of the variation of ZrC x properties with stoichiometry have been summarized by Snead, et al. [18]. Because the properties change with stoichiometry, it is necessary to consider the composition when performing any experiments with this material. Point defects in ZrC have been studied by Kim, et al. [16] who used ab initio calculations to determine point defect formation energy. A summary of the point defect energies is shown in Table 1.2. It was found that the formation energy of C vacancies and interstitials are much lower energy than that of Zr vacancies and interstitials. The stable C interstitial structure was found to be a C-C-C trimer in the [101] direction, and the stable Zr interstitial structure was found to be a centered double tetrahedron. It should be noted that the formation energy of the C interstitial located in the center of a C tetrahedron was found to be only be 0.26 ev/defect higher than the trimer. These interstitial configurations are shown in Figure 1.4. In addition, the anti-site defects, a Zr atom in a C lattice position and vice versa (Zr C and C Zr ), have the highest defect formation energies. As a result, the anti-site defects would be energetically unfavorable when compared to the individual point defects. A tendency for the lower formation energy defects to form during irradiation, either directly or as a result of defect-defect reaction, should be expected. Table 1.2. Formation energy [ev/defect] of point defects in ZrC [16]. Va Zr Va C I Zr I C Zr C C Zr

19 7 Figure 1.4. Interstitial defect configurations reported by Kim, et al. [16]. 1.3 ZrC for TRISO The most basic requirement for the use of ZrC in TRISO fuels is the ability to consistently manufacture a high quality coating. A chemical vapor deposition process (CVD) utilizing zirconium bromide was successfully used to coat pyrolytic carbon with ZrC in a recent paper by Ueta, et al. [4]. Although problems were initially encountered as circumferential stripes were observed in the ZrC due to free carbon, a uniform ZrC structure was achieved through better temperature control at 1400 C. Similarly, but using a chloride CVD process, Chernikov and Kosukhin [19] were able to produce pyrolytic carbon, ZrC, and two-phase mixtures of the two by systematically controlling the process temperature and gas composition. Heating the ZrC up to 2600 C showed recrystallization, but no cracks or pores. Zhao, et al. [20] performed extensive analysis of a ZrC coating deposited on pyrolytic carbon coated particles using a chloride CVD

20 8 method. They found equiaxed single phase grains with average diameter of about 20 to 50 nm and nearly isotropic texture. In addition, the ZrC formed was found to be close to stoichiometric. These studies demonstrate the feasibility of producing high quality ZrC coatings, but it has yet to be shown that ZrC-TRISO particles can be manufactured on the large scale that would be required for VHTR fuel production. As stated previously, a primary function of the ZrC coating is to retain fission products. With respect to this, some investigations into the diffusion of metallic fission products in ZrC have been performed. Stark [21] studied the diffusion of cesium in ZrC coated graphite spheres by exposing them to Cs vapor. Neutron activation analysis was then used to measure the cesium concentration and calculate the diffusion coefficient. The experiments were performed at temperatures ranging from 1485 to 1896 K and the diffusion coefficient was calculated to be within the range of to m 2 /s, which is similar to the diffusion coefficient of Cs in SiC. Ogawa and Ikawa [22] also studied the diffusion of Ru, Sr, and Ba in ZrC using Sr soaking and post-activation annealing methods. The results were promising in that they found diffusion coefficients that were comparable to those in SiC. The findings of these investigations indicate that fission product retention by ZrC may be on a par with SiC. In more direct tests of the feasibility of ZrC coatings, ZrC-TRISO particles have been manufactured and irradiated in-reactor. Ogawa, et al. [23] performed irradiation and post-irradiation heating of ZrC-TRISO fuel particles. They found minimal in-reactor release of fission products which they ascribed to contamination of the graphite matrix. Examination of the particles after irradiation showed no degradation of the ZrC by palladium as was found with SiC. Although the particles maintained their integrity up

21 9 to 2400 C, one failure was observed among approximately 100 particles after holding at that temperature for 6000 seconds. Minato, et al. [24] also irradiated and performed post-irradiation heating of ZrC-TRISO particles. The particles were irradiated to 1.5% fissions per initial metal atom (FIMA) at 900 C, and then heated to either 1800 C for 2000 hours or 2000 C for 100 hours. They found some degradation of the ZrC layer but no through-coating failure. At 1800 C, the diffusion coefficient for cesium in ZrC was calculated to be two orders of magnitude lower than in SiC, and the diffusion coefficient for Ru was comparable. Some release of Cs was observed during the heating to 2000 C. In another article by Minato et al.[25], they found that the release of cesium was caused by the deterioration of ZrC by reaction with CO after failure of the inner pyrolytic carbon layer. They speculated that this may be avoided by ensuring a high quality IPyC layer. Overall, as seen from the preceding, the performance of irradiation tests on ZrC-TRISO fuel was promising. One hypothetical accident condition in a VHTR involves a break in the coolant loop and a significant ingress of air into the reactor vessel. In such a case, high temperature oxidation of ZrC may be of concern. Kuriakose and Margrave [26] oxidized ZrC at C and found linear oxidation kinetics and a non-protective, porous film oxide. In that study, a sample was also heated in an oxygen atmosphere and was found to begin to crumble at 700 C, before finally breaking up at 850 C. In another experiment, Opeka, et al. [27] continuously heated ZrC at a rate of 20 C per minute and found the sample to be completely oxidized by 700 C. Shimada, et al. [28] exposed ZrC single crystals to 2.6 kpa O pressure at 500, 550, and 600 C. They found the initial oxide to be an amorphous ZrO 2 that grew parabolically to 2-3 micrometers. The oxidation kinetics

22 then became linear; the amorphous layer remained next to the ZrC and remained at a 10 constant thickness, but the oxide far from the ZrC/oxide interface became cubic ZrO 2 crystallites. Bellucci, et al. [29] also oxidized single crystals of ZrC at 1 bar pressure and C and found the oxygen consumption with time to be linear. Experimental characterization of the oxide using x-ray diffraction (XRD) and Raman techniques identified both cubic and monoclinic ZrO 2 in the oxide. These experiments indicate that ZrC is susceptible to high temperature oxidation, and this would need to be considered in the application of ZrC in a VHTR with respect to accident conditions. 1.4 Radiation Damage in ZrC An important consideration with respect to the use of ZrC in TRISO fuel is the response of the material to irradiation. Both the changes in the macroscopic properties of ZrC due to irradiation and their impact on the performance of the fuel particle are controlled by the microstructural changes that occur during irradiation. As a result, fundamental understanding of radiation damage is essential to the complete understanding of fuel performance. The microstructural evolution of ZrC under irradiation has been investigated in a number of studies, and the experiments and their results are described below. Gan, et al. [30] conducted in situ irradiations on ZrC 1.01 at the intermediate voltage electron microscope (IVEM). The IVEM is a transmission electron microscope (TEM) that has been interfaced with an ion accelerator in order to perform irradiations while simultaneously imaging the microstructure. The samples were irradiated with 1 MeV Kr ions to 10 and 30 dpa at 27 C and 10 and 70 dpa at 800 C with a dose rate of

23 11 approximately dpa/s. At 27 C, a high density of black dot damage was observed, and at 800 C a lower density of small dislocation segments was observed. There was a general lattice constant increase of approximately 0.6% after irradiation, except for 70 dpa at 800 C where an increase of 7% was recorded. During these irradiations, ring intensity formed in the diffraction patterns and is consistent with the formation of an FCC phase with lattice constant approximately 8% larger than ZrC. A study using 2.6 MeV protons to irradiate bulk ZrC 1.01 was carried out by Yang, et al. [31] The experiment was conducted at 800 C with a dose rate of dpa/s to total doses of 0.7 and 1.5 dpa. A high density of small loops was observed, as shown in Figure 1.5. The average diameter increased from 4.3 nm at 0.7 dpa to 5.8 nm at 1.5 dpa. The density similarly increased from m 3 at 0.7 dpa to m 3. The rel-rod technique was used to identify faulted Frank loops. A defect denuded zone was observed near grain boundaries. Lattice expansion remained less than 0.1%. An increase in hardness and fracture toughness was also measured. Figure 1.5. TEM micrographs of the irradiated microstructure of ZrC observed at 0.7 and 1.5 dpa by Yang, et al. [31].

24 12 Gosset, et al. [32] irradiated ZrC using 4 MeV Au ions. The work was performed at room temperature to fluences ranging from to ions/cm 2. X-ray diffraction measurements showed swelling with strains saturating at a value of at ions/cm 2, corresponding to a few dpa. TEM examination showed a high density of small faulted loops at ions/cm 2 evolving into a dense dislocation network at ions/cm 2. In addition, diffraction rings were observed and attributed to nanometer-sized crystals of quadratic zirconia. Neutron irradiations were performed on ZrC 0.87 by Snead, et al. [18] using the High Flux Isotope Reactor. The fast neutron (E >0.1 MeV) flux was neutrons/m 2 s, and total fast neutron fluence ranged from neutrons/m 2. Irradiation temperatures ranged from C. At 635 C, a dense network of unidentified loops was observed. By 1023 C, the microstructure had coarsened and was composed of Frank loops. Increasing the temperature further to 1496 C showed further coarsening of the microstructure and a transition from Frank loops to large prismatic loops. TEM micrographs of the microstructure are shown in Figure 1.6. Measurements showed an increase in hardness and fracture toughness and a slight decrease in elastic modulus and thermal conductivity. The swelling strain remained less than 0.2%. As can be summarized from the above, the work in the area of fundamental radiation damage in ZrC is incomplete. The experiments were limited in the range of temperatures studied; most focused on one or two temperatures, and most were performed at high temperature. There has been no systematic investigation into the effect of temperature on the accumulation of radiation damage in ZrC, and no irradiations have been performed at cryogenic temperatures where athermal reactions can be studied. In

25 13 Figure 1.6. TEM micrographs of the neutron irradiated microstructure of ZrC observed by Snead, et al. [18] at various temperatures. addition, the effect of stoichiometry on the irradiation response of ZrC x has not been investigated in any experiment. It is known that the properties of ZrC x are sensitive to the stoichiometry, and this may affect the irradiation response. Finally, most of the experiments are limited in the number of dose points observed. Thus a great deal can be derived from studying the detailed kinetics of the irradiation damage development by performing irradiation in situ using the IVEM. 1.5 Goals of the Study The goals of this research are to gain a greater understanding of the fundamental irradiation response of ZrC. To gain this understanding, irradiations are performed on ZrC using ions and the microstructure is simultaneously investigated using TEM. A wide range of irradiation temperatures is used, from cryogenic temperatures at which defects should have minimal or no mobility to high temperatures where the agglomeration of defects into larger clusters can be observed. The irradiation temperature is systematically varied so that its detailed effect on defect kinetics can be precisely determined.

26 14 Samples of different, known stoichiometry are prepared and irradiated to determine how the stoichiometry affects the microstructural evolution under irradiation. Additionally, the experiments are conducted in situ so that the microstructure can be continuously followed, allowing for the direct observation of defect accumulation mechanisms as they occur. The defect size distribution and density is determined as functions of temperature and dose. Finally, the effects of existing microstructure, such as dislocations and the sample surface are investigated.

27 15 Chapter 2 Experiment The details of the experiments are explained in this chapter. The source of the samples is provided, including the manufacturing process and sample preparation techniques. The unirradiated material is characterized as a reference point for irradiation experiments. The irradiation procedures at the intermediate voltage electron microscope (IVEM) are detailed and the process used for experimental analysis is outlined. Finally, the calculation of dpa equivalent dose is provided. 2.1 Material The samples of zirconium carbide used for this research were purchased from Applied Physics Technologies, Inc. Polycrystalline samples are produced by an arc, floating zone refining process that is an extension of the work published by Mackie and Hinrichs [33]. The starting stock is a hot-pressed or sintered material made from >99.5% purity powder. The stoichiometry of the finished material is controlled by adjusting the starting powder concentration. The sintered material is then cut into strips to be placed in the zone refiner. Once the sample is mounted, the chamber is evacuated and backfilled with argon. Two tungsten electrodes placed on either side of the material create a molten zone as current passes between them. The melted region is held in place by the surface tension with the solid material above and below it. The electrodes are then

28 slowly moved along the sample; the previously melted region re-solidifies again after the passage of the electrodes. Polycrystalline samples are made by increasing the speed of 16 the electrodes while excluding a seed crystal. When finished, the sample is cut and ground to the desired length and diameter. The purchased samples have five nominal stoichiometries of ZrC 0.8, ZrC 0.9, ZrC 1.0, ZrC 1.1, and ZrC 1.2. The compositions were verified by the ATI Wah Chang Analytical Laboratory using a combustion-ir method. The results are presented in Table 2.1. The as-received samples are cylindrical rods with a diameter of 3 mm and an average length of 3.2±0.2 cm. This particular geometry is convenient for preparation of 3 mm diameter transmission electron microscope (TEM) disks. Figure 2.1 shows a photograph of an as-received sample after being cut to approximately half its original length. Table 2.1. Measured stoichiometry for samples of ZrC x. Nominal Stoichiometry ZrC 0.8 ZrC 0.9 ZrC 1.0 ZrC 1.1 ZrC Stoichiometry Measurements Sample Preparation The initial characterization of the samples was performed using optical microscopy and transmission electron microscopy. The basic sample used for each technique was a 600 µm thick disc cut from the as-received rods using a low speed saw with a 6 mil thick

29 17 Figure 2.1. Photograph of as-received ZrC rod. metal-bonded diamond wafering blade. For optical microscopy, the samples were then ground and polished using successive steps of 400, 600, 800 and 1200 grit SiC abrasive discs. Final polishing was performed using 1 µm diamond polishing solution. Grains were revealed in the sub-stoichiometric samples by electrolytic etching with a procedure similar to that used to produce TEM disks, which is described next. Samples were prepared for TEM by sectioning the sample rods into discs as previously discussed, followed by grinding the samples down to approximately 200 µm thickness using 400 and 600 grit SiC abrasive discs, and finally followed by a final surface finish using an 800 grit SiC abrasive disc on both sides. The discs were then prepared for TEM examination by an electrolytic polishing method using a Struers TenuPol-5. This is a jet polishing method in which the center of the samples are thinned to electron transparency. A solution of 7.5% perchloric acid and 92.5% methanol cooled to 243 K was used for the electropolishing. Other parameters for the TenuPol-5 included a voltage of 25 V, light shutoff value of 60, and pump flow rate of 26. Immediately after electropolishing, the samples were removed and rinsed with methanol and ethanol. They were finally stored

30 18 in pure ethanol to prevent possible surface degradation. It should be noted that this procedure produces excellent TEM samples of single phase, sub-stoichiometric ZrC; it was less successful for samples containing the free carbon phase, discussed later, because of different removal rates of the two phases. 2.3 Initial Characterization Optical microscopy was performed on each of the stoichiometries, and low magnification images of the general microstructure are shown in Figure 2.2. The substoichiometric samples of ZrC 0.8 and ZrC 0.9 are homogeneous with a single phase. The grains of these samples, as revealed by the electrolytically etched sample shown in Figure 2.3, are quite large with dimensions consistently greater than 100 µm. The other samples, ZrC 1.0, ZrC 1.1 and ZrC 1.2, show a two-phase structure. As ZrC x approaches stoichiometry x = 1, a free carbon (graphite) phase is formed in the material. As the carbon fraction increases from ZrC 1.0 to ZrC 1.1 to ZrC 1.2, the fraction of the graphite phase also increases. In Figure 2.4, higher magnification images show the fine microstructure of the two phase samples. The carbon phase, showing dark contrast in the images, forms in the grain boundaries of ZrC and, in larger two-phase regions, forms a layered structure of ZrC and graphite. The observed microstructures, including the phases present, agree well with the phase diagram from Figure 1.3. Initial TEM characterization of the unirradiated samples of ZrC 0.8 and ZrC 0.9 was performed to determine the base microstructure before irradiation. The microscopy revealed a large electron transparent region produced by the electropolishing preparation. Because of the large grains of the samples, there was only a single grain present in the

31 Figure 2.2. Optical microscope images showing the general microstructure of ZrC x. 19

32 20 Figure 2.3. Optical microscope image showing the grains of ZrC 0.9. observable region for nearly every sample produced. The microstructure was very clean of defects; only a few pre-existing dislocations were present. A relatively low magnification micrograph is presented in Figure 2.5 which shows the large electron transparent region as well as examples of two pre-existing dislocations; this microstructure is typical of all ZrC samples produced with good electropolishing. 2.4 IVEM Irradiation Details In situ ion irradiation experiments were conducted at the IVEM-Tandem Facility at Argonne National Laboratory [34]. This facility incorporates a TEM that has been modified and interfaced with an ion accelerator; the major advantage of this setup is the ability to simultaneously image, including recording video of, the sample as it is being

33 Figure 2.4. Optical microscope images showing the two-phase structure of ZrC

34 22 Figure 2.5. A 3g dark-field TEM micrograph showing the unniradiated microstructure of ZrC 0.9. Pre-existing dislocations are marked by small arrows. irradiated by ions. It is possible to continuously follow the microstructure as it changes during irradiation. The IVEM is a Hitachi H-9000NAR TEM. This instrument uses a LaB 6 emitter and can operate at accelerating voltages ranging from kev. The spherical aberration coefficient is 2.8 mm, and the point resolution at 300 kev is 0.25 nm. The vacuum in the specimen chamber is less than Torr. A variety of sample holders are available, including double-tilt heating and cooling, tensile testing, and ambient temperature double-tilt and rotate. The microscope is interfaced with two NEC accelerators: a 650 kv ion implanter and a 2 MV tandem Van de Graaf. The ion beam is tilted 30 from the electron column, and the ion beam diameter at the sample is approximately 1.5 mm. The ion beam current is measured close to the sample using a skim Faraday cup and a movable Faraday cup. Figure 2.6 shows a photograph of the IVEM and a sketch of the ion beam and TEM interface in the microscope.

35 23 Figure 2.6. A photograph of the IVEM (left) and a sketch of the ion beam and TEM interface[34] (right). The irradiations were carried out using 1 MeV Kr ++ ions with a flux of ions/cm 2 /s. Double tilt heating and cooling specimen holders were used during these experiments. The heating holder is made by Gatan and can operate up to 1173 K. The heating is performed electrically and is adjusted by hand. The cooling holder is cooled by liquid helium and can reach temperatures as low as 15 K. Experiments were performed at temperatures ranging from 20 K to 1073 K. It was decided, considering the limited availability of the facility, to irradiate only ZrC 0.8 and ZrC 0.9. It was possible to consistently create excellent TEM samples for these stoichiometries, and the presence of a single phase reduced the number of variables and allowed for a focus on the fundamental irradiation response of ZrC x. The list of all irradiations performed is given in Table 2.2. For these experiments, a 300 kev TEM accelerating voltage was initially used for the irradiations performed at 20 K and 50 K. However, due to the possibility of producing

36 displacement damage by electrons, the accelerating voltage was decreased to 200 kev for the experiments conducted at every other temperature. 24 Table 2.2. List of experiments conducted at the IVEM in this work. Material T [K] Max. Dose [dpa] ZrC ZrC The irradiated microstructure was continuously observed throughout the irradiations while recording digital video. At specific dose points, the irradiations were paused to adjust instrument settings and acquire high quality images. A combination of bright field imaging, dark field imaging, and diffraction patterns were used to characterize the irradiated microstructure. Bright field and dark field imaging are techniques where diffraction contrast is created due to the strain field around the core of dislocations. An aperture is placed in the back focal plane to choose the desired diffracted beam for imaging. Selected area diffraction is a technique in which an aperture is place in the image plane to only allow information from a particular area of the sample to be collected. Then, the back focal plane is focused on the imaging device and a diffraction pattern

37 is recorded. This pattern can provide crystallographic information about the selected region of interest Experimental Analysis Dark-field TEM images were analyzed to determine average defect density and diameter as functions of dose and irradiation temperature. Defects were manually and individually counted using the program ImageJ. When defects were non-circular, the longest dimension was measured as the defect diameter. A square area with an edge length of 215 nm, or a total area of 4.63x10 14 m 2, was analyzed for each dose point. In order to calculate the defect density, the total volume of the analyzed areas needed to be determined. Therefore, the local thickness of the sample was needed to calculate the defect density. The sample thickness was estimated using thickness fringes. The direct beam intensity for a two-beam approximation at the exact Bragg condition varies with sample thickness as ( ) φ 2 2 πt 0 = 1 sin ξ g (2.1) where φ 0 2 is the intensity of the direct beam, t is the sample thickness and ξg is the extinction distance. The extinction distance can be calculated using ξ g = πv c cosθ B λf g (2.2) where V c is the volume of the unit cell, θ B is the Bragg angle, λ is the electron wavelength, and F g is the structure factor at the Bragg condition. The structure factor for ZrC is

38 written as F B = 4(f Zr + f C ) if h, k, and l are all even 4(f Zr f C ) if h, k, and l are all odd 0 otherwise 26 (2.3) where f Zr and f C are the atomic scattering amplitudes for Zr and C and h, k, and l are the indices of planes [35]. Using values of 4.7 Å for the lattice parameter, 200 kev for the electron energy, and the atomic scattering factors tabulated by Colliex et al. [36], the extinction distances for the three diffraction conditions used in these experiments were calculated and are shown in Table 2.3. These values were then be used to estimate the local thickness of the samples when imaged using a two-beam bright-field condition. An example is shown in Figure 2.7, where a sample of ZrC 0.9 was imaged using a g = 220 two-beam brightfield condition; the thickness, based on thickness fringes and the calculated extinction distance, is marked. Table 2.3. Calculated extinction distances for ZrC. Diffraction Condition, g Extinction Distance, ξ g [nm] Additionally, the extinction distance for the g = 220 condition was verified using convergent beam electron diffraction (CBED) as outlined by Kelly, et al. [37]. This technique uses the intensity oscillations observed in convergent beam diffraction patterns

39 27 Figure 2.7. Example of local thickness in ZrC 0.9 determined by thickness fringes in a g = 220 two-beam bright-field condition.

40 28 which are governed by the equation ( si n i ) 2 ( ) ( ) = n i ξ g ( ) 1 2 (2.4) t where n i is a whole number for the i-th intensity minimum (numbered in order from inside to outside), ξ g is the extinction distance, t is the sample thickness, and s i is defined as s i = λ ( ) θi d 2 2θ d (2.5) where λ is the electron wavelength, d is the spacing of the diffracting planes, and θ i and 2θ d are as defined in the example CBED image shown in Figure 2.8. After measuring the intensity oscillations, the values of (s i /n i ) 2 are plotted against (1/n i ) 2, and then the slope is equal to (1/ξ g ) 2. This procedure was completed for g = 220 at three different thicknesses in ZrC 0.9, and the results are shown in Figure 2.9. The extinction distance was measured to be 56.6 nm, which is approximately 5% greater than the calculated value. The analysis was performed using a 200 kev TEM acceleration voltage. Figure 2.8. CBED pattern recorded with g = 220 in ZrC 0.9. The direct beam is on the left and the diffracted beam is on the right.

41 29 Figure 2.9. CBED analysis performed at three thicknesses to determine the extinction distance for g = 220 in ZrC Irradiation Dose Another measure of the irradiation dose given to a material is displacements per atom (dpa). This is a measurement of the the average number of times an atom has been displaced from its lattice site during the irradiation. This quantity cannot be directly measured, but is instead calculated based upon the experimental parameters, including the characteristics of the radiation and the irradiated material. One important quantity is the threshold displacement energy which is a measure of the minimum energy that needs to be imparted to an atom to displace it from its lattice position and produce a stable Frenkel pair, or interstitial and vacancy pair. For experiments where the radiation

42 30 damage in the form of lattice defects created in the material is of concern, the displacements per atom is a more relevant measure of radiation damage than the ion fluence. It is also more straightforward to compare between different experiments. The Stopping and Range of Ions in Matter (SRIM)[38] is a freely available software package that can be used to calculate equivalent displacements per atom from ion irradiation. SRIM is a Monte Carlo simulation that, by using a binary collision approximation, follows many individual ions as they impinge on a material. While making use of a universal screening function, flight distances and collision details are calculated and used to find the spatial distribution of displacements in the material. An example of the input parameters for these calculations is given in Table 2.4; in this case, it is found that the average number of displacements per ion after trials is 1485 ± 29 (95% confidence interval). In other terms, 1 dpa corresponds approximately to ions/cm 2. Given the ion flux used in these experiments, the dose rate was calculated to be dpa/s. The full cascade option was chosen because there is no explicit formulation for materials composed of multiple elements in the Kinchin-Pease equation [39]. Table 2.4. Input parameters for SRIM simulation. Calculation Type full cascade Ion Species krypton Ion Energy 1000 kev Target Width 1000 Å Target Density 6.5 g/cm 3 Target Composition ZrC 0.9 Displacement Energy Zr = 35 ev [40] C = 24 ev [40] Number of Ions 10000

43 31 The damage distribution profile for this simulation is provided in Figure From the plot, it is observed that the number of carbon displacements is 80% of the number of zirconium displacements; when taking into account the stoichiometry, the damage per atom for C is 90% of that for Zr; the damage to each element is at similar levels. Over 97% of the ions pass completely through the sample, meaning that there is little implantation of Kr. In addition, information extracted from the COLLISON.TXT file was used to produce the pka spectrum in Figure The average primary knock-on atom (pka), or the atoms that directly collided with the irradiating species, energy was 3.64 kev, with average pka energies for the individual elements being 4.83 kev for Zr and 1.75 kev for C. The pka spectrum for ion irradiations is softer than that from neutron irradiations [15]. When the same simulation is performed for ZrC 0.8, the total damage rate per atom increases slightly, by less than 5%. Figure ZrC 0.9. SRIM calculated damage profile for 1 MeV Kr ions into 100 nm thick

44 32 Figure ions. Integral pka spectrum as calculated by SRIM for ZrC 0.9 and 1 MeV Kr

45 33 Chapter 3 Characterization of Irradiated Microstructure This chapter presents and discusses the observations made during the in situ irradiation experiments. First, a general overview will be given wherein the dose evolution of the microstructure is followed for selected irradiation temperatures. Then, more specific observations are detailed. Quantitative analysis of average defect diameter and density is performed. The rings formed in the diffraction patterns during irradiation are indexed. The possible habit planes of dislocation loops are investigated by comparing with theoretical projections of some loop orientations. 3.1 Dose Dependence of Damage Accumulation In this section, the basic microstructural evolution is described for select irradiation temperatures, starting at cryogenic temperatures of 20 and 50 K and moving to the highest temperature, 1073 K. It is shown that the irradiated microstructure is dependent on temperature, such that irradiations at a high temperature result in a coarser microstructure with larger defects. The results for ZrC 0.8 and ZrC 0.9 are presented together for each temperature to better allow for comparison between the two. When indicated in a dark-field image, the diffraction condition represents the (g,3g) semi-weak beam imaging condition.

46 Microstructure Evolution at 50 K and Below Irradiations were performed at cryogenic temperatures of 20 K and 50 K up to doses of 5 and 10 dpa, respectively. At these temperatures, evidence of damage in the form of small, black-dot defects was first observed at doses as low as 0.4 dpa. A progression of the damage accumulation in ZrC 0.9 at 50 K is shown in Figure 3.1. The density of damage increased quickly with dose up to approximately 2 to 3 dpa, at which point the rate of damage accumulation slowed. Figure 3.2 shows the microstructure of ZrC 0.9 resulting from a dose of 2 dpa at 50 K. A high density of black-dot damage is observed. The micrograph in Figure 3.2 is centered on a pre-existing dislocation, indicated by an arrow. There is no apparent defect-denuded zone surrounding the dislocation, such as previously observed near grain boundaries after irradiation of ZrC with protons to a dose of 1.5 dpa at 1073 K [31]. The microstructure of ZrC 0.9 after irradiation to 10 dpa at 50 K is shown in Figure 3.3. The selected area diffraction pattern shows the development of clearly defined diffraction rings, indicated by the arrows. The extra sets of rings that appear around diffracted spots are caused by double diffraction. A dark-field micrograph imaged using the diffracted intensity from the rings reveals that they are associated with the larger sized dark spots of the corresponding bright-field micrograph. This indicates that a different, unidentified phase forms under irradiation; attempts have been made to index the diffracted rings and will be discussed in a later section. The dark-field micrograph imaged using the diffracted intensity from the ZrC matrix shows the presence of small, black-dot defects (Figure 3.3b).

47 35 The irradiation of ZrC 0.9 at 20 K produced similar results to those observed at 50 K, i.e. a high density of small, black-dot type damage. No amorphization was observed, even after doses of 10 dpa at 50 K and 5 dpa at 20 K. In contrast with other in situ observations in steels [41], the defect clusters that formed were immobile during irradiation, and their density simply increased to saturation rather than exhibiting a dynamic equilibrium.

48 Figure 3.1. A series of bright-field micrographs showing ZrC 0.9 irradiated to 1, 3 and 5 dpa. 36

49 37 Figure 3.2. Dark-field micrograph of ZrC 0.9 irradiated to 2 dpa at 50 K centered on a pre-existing dislocation (marked by arrow). No defect-denuded zone is seen.

50 38 Figure 3.3. Microstructure of ZrC 0.9 irradiated to 10 dpa at 50 K. (a): bright-field micrograph. (b): dark-field micrograph using diffracted intensity from ZrC matrix. (c): dark-field micrograph using diffracted intensity from the brightest ring down and to the left from the beam stop. (d): selected area diffraction pattern.

51 Microstructure Evolution at 300 K The microstructural evolution of ZrC 0.9 irradiated at room temperature (300 K) with increasing dose is shown in Figures 3.4 and 3.5; similarly, Figures 3.6 and 3.7 show the microstructural evolution of ZrC 0.8 during irradiation at room temperature. The bright line in the upper-left corner of the images of ZrC 0.8 is due to a pre-existing dislocation that was used to identify the area of observation. Because the microstructural evolution in both materials was similar, they will be discussed together here. The first indicator of irradiation effects is a mottling of the background, as compared to the unirradiated condition, seen in the micrographs at 0.1 dpa and 0.3 dpa. The exact cause of this mottling is unknown, but may be due to an accumulation of sub-visible lattice defects or alternatively due to a roughening of the surface due to sputtering. At this temperature, the first definitive observation of irradiation-induced defects occurs at 0.5 dpa for both compositions. It is possible that some defects have appeared by 0.3 dpa, but the contrast is low and it is difficult to re-identify the same defects at higher doses. The defects seen at 0.5 dpa appear as black-dot damage and measure approximately 2 to 4 nm in diameter. As the irradiation dose increases, both the number of defects and the defect diameter tend to increase. There are no observations of dynamic defect interaction; existing defects can be followed throughout the experiment and defect growth occurs slowly throughout the irradiation. By 5 dpa, there is a high density of black-dot damage that appears to have saturated or is well into the process of saturating. The size of the defects ranges from the 2-4 nm observed at lower doses up to 5-7 nm in diameter.

52 40 There is no direct visual evidence to indicate the nature of the defects formed at this temperature, but the defects are hypothesized to be small, unresolved dislocation loops. No changes to the diffraction pattern, including rings or streaking, were observed for the irradiations conducted at 300 K or 473 K.

53 Figure 3.4. Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 300 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa. 41

54 Figure 3.5. Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 300 K showing the microstructure at 3, 4, and 5 dpa. 42

55 Figure 3.6. Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 300 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa. 43

56 Figure 3.7. Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 300 K showing the microstructure at 3, 4, and 5 dpa. 44

57 Microstructure Evolution at 873 K The dose evolution of the microstructure during the irradiation of ZrC 0.9 to 0.5 dpa at 873 K is shown in Figure 3.8. Similarly to the irradiation at 300 K, a mottling of the background is observed at very low doses (see the dark-field image for 0.1 dpa). However, in contrast to the irradiation at 300 K in which visible defects were not found until 0.5 dpa, a number of visible defects are already visible by 0.3 dpa at 873 K. These initial defects are black-dot damage. It should be noted that the bright mottled areas at 0.1 dpa appear to indicate the future locations of some visible defects at 0.3 dpa, meaning that the mottled contrast could be due to small unresolved defects. Both defect size and density increase as the dose increases to 0.5 dpa, but the black-dot damage morphology remains. Figure 3.9 shows the continued evolution of the ZrC 0.9 microstructure irradiated to 1-4 dpa at 873 K. The size and density of defects continues to increase with dose. The density appears to trend toward saturation at these doses. The defect diameters have increased from 2-5 nm at 0.3 dpa up to a maximum of approximately 20 nm at 4 dpa. Some loop-like defects begin to form around 2 dpa and become well defined by 4 dpa. No migration of the visible defect clusters is observed at 873 K. Defects slowly form and then remain in the microstructure throughout the experiment, growing slowly with dose. The microstructures observed during the companion experiment in which ZrC 0.8 was irradiated at 873 K are shown in Figures 3.10 and The observed microstructures are similar to those observed in ZrC 0.9. The initial defects are formed by 0.3 dpa, and

58 46 the size and density increase with dose. Loop-like structures also form by 2 dpa and continue to grow as the dose progresses to 5 dpa. At 873 K, the large defect diameters, coupled with their high density, make it difficult to characterize individual loops because of significant overlap in the projected image. A ring pattern develops in the diffraction pattern during irradiation, as shown in Figure 3.12, and is discussed in a later section. There is a concentration of diffracted intensity to specific points along the rings which may indicate a preferred orientation. The diffraction pattern also shows streaking from {111} to {220} type diffraction spots, which could be caused by an array of planar defects or dislocations [35]. Note that the two vertical streaks passing through the two brightest diffraction spots are artifacts.

59 Figure 3.8. Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 873 K showing the microstructure at 0.05, 0.1, 0.3, and 0.5 dpa. 47

60 Figure 3.9. Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 873 K showing the microstructure at 1, 2, 3, and 4 dpa. 48

61 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 873 K showing the microstructure at 0, 0.3, 0.5, and 1 dpa. 49

62 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 873 K showing the microstructure at 2, 3, 4, and 5 dpa. 50

63 Figure A diffraction pattern taken from ZrC 0.9 irradiated to 5 dpa at 873 K. 51

64 Microstructure Evolution at 1073 K Figure 3.13 shows the microstructure evolution of ZrC 0.9 when irradiated at 1073 K to doses up to 2 dpa. Damage is initially observed at 0.3 dpa and consist of black-dot defects. The defects increase in size and density up to 1 dpa; at this dose, a few small loops are present. After doubling the dose to 2 dpa, some coalescence of smaller defects into loops up to approximately 30 nm in diameter is observed. There is no apparent long-range diffusion of visible defects. However, after higher densities are achieved at approximately 1 dpa, interaction of closely spaced defects occurs, likely driven by interactions of defect strain fields. In addition, the slow growth of defects observed at lower temperatures continues to occur. The microstructure of ZrC 0.9 at doses ranging from 3-10 dpa is shown in Figure The microstructure continues to coarsen with dose up to 4 to 6 dpa. After that point, there is little visible change in the defected microstructure. The microstructure at these doses is difficult to characterize due to a high density of overlapping defects. The appearance is that of a tangled dislocation structure, but could actually result from the projection of overlapping dislocation loops. At 8 and 10 dpa, the images start to become washed out because the objective aperture admits some of the diffraction ring intensity, which becomes stronger with dose. Figures 3.15 and 3.16 show the irradiated microstructure of ZrC 0.8 up to 5 dpa at 1073 K. Once again, the dose evolution of the irradiated microstructure of ZrC 0.8 is very similar to ZrC 0.9. Initial defects are found at 0.3 dpa as black-dot damage where density and size increase with dose. At 1 dpa and beyond, larger dislocation loops begin to form

65 53 and a general coarsening of the microstructure is observed to 5 dpa. Some coalescence of the small defects into larger dislocation loops is observed. The final microstructure at 5 dpa consists of either a tangled dislocation structure or a high density of overlapping dislocation loops, similar to that of the other stoichiometry. A series of images were captured from the video recorded between 1 and 2 dpa during the irradiation of ZrC 0.9 at 1073 K. Figure 3.17 shows these images in steps of 0.1 dpa. The images are centered on the location of a dislocation loop that is not present at 1 dpa but develops by 2 dpa. Sketches of the loop shape are provided to the right of the loop when observed. Initially, only black-dot damage is observed at 1 dpa. It is difficult to characterize the initial formation of the loop, but it is first observed by 1.4 dpa and has a non-circular shape, as shown. The loop expands and becomes more circular by 2 dpa. The inside of the loop, where there existed several black-dot type defects before, appears devoid of defects which may indicate coalescence of defects during the formation of the loop.

66 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 1073 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2 dpa. 54

67 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.9 at 1073 K showing the microstructure at 3, 4, 6, 8, and 10 dpa. 55

68 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 1073 K showing the microstructure at 0, 0.1, 0.3, 0.5, 1, and 2. 56

69 Figure Series of dark-field TEM images acquired during the irradiation of ZrC 0.8 at 1073 K showing the microstructure at 3, 4, and 5 dpa. 57

70 58 Figure Series of dark-field TEM images captured from the video taken during the irradiation of ZrC 0.9 at 1073 K between 1 and 2 dpa showing the development of a dislocation loop. The images are centered on the dislocation loop location, and sketches of the loop shape, when discernible, are inset to the right.

71 Defect Measurement and Counting The microstructure developed during irradiation was further characterized by quantitative measurement of the average defect diameter and defect density. This was done by using the ImageJ software package to individually measure each defect from a selected square region with side length of 215 nm. For non-circular defects, the defect diameter was measured along the longest axis. Once all of the defects were measured, the defect size distribution, average defect diameter, and number of defects could be calculated. It should be noted that due to the very high density of defects resulting in overlapping of defect projections, the analysis could only be performed up to 3 dpa. An example size distribution is provided in Figure 3.18 and shows the distribution of defects for ZrC 0.9 irradiated to 1 dpa at 673 K. The distribution shape is typical of all measurements and can be described as either a Poisson or normal distribution. At higher irradiation temperatures, the average shifts right to larger defect diameters and the distribution broadens. The size distribution clearly shows the minimum defect resolution of approximately 2 nm (defects < 2 nm are invisible). The average defect diameter and defect density in ZrC 0.9 are plotted as functions of dose in Figure 3.19 for 673 K, 873 K, and 1073K. The standard errors of the measurements are marked by the error bars. The defect density was calculated by estimating the specimen thickness using thickness fringes as discussed in Section 2.5. The average defect diameter increases with dose and irradiation temperature. Over the range from 1 to 3 dpa, the average diameter reaches 4 to 5 nm at 673 K and 873 K and 6 to 8 nm at 1073 K. The rate of increase in average diameter up to 1 dpa is greater than from 1

72 60 Figure Defect size distribution for ZrC 0.9 irradiated to 1 dpa at 673 K. to 2 dpa or 2 to 3 dpa, which may be indicative of saturation of defect size. The defect density was on the order of to per cubic meter. For the irradiations at 673 K and 873 K, the defect density increases with increasing dose and trends toward saturation at higher doses; however, saturation was not reached by 3 dpa in either case. For the irradiation at 1073 K, the defect density increases up to 1 dpa, and then decreases slowly until 3 dpa; this decrease in density may be due to coalescence of smaller defects into larger defects. 3.3 Temperature Dependence of Defect Morphology It is interesting to highlight the effect of temperature on defect morphology. Shown in Figure 3.20 is the irradiated microstructure observed in ZrC 0.9 irradiated to 1 dpa at temperatures ranging from 300 K to 1073 K. In all cases, the overall microstructure observed is small, black-dot damage. The image contrast improves continually with increasing irradiation temperature. Defects are more clearly seen, which may indicate

73 61 Figure Quantitative measurements of the average defect diameter and defect density for ZrC 0.9 irradiated at 673 K, 873 K, and 1073 K. either larger defects at high temperature, a greater number of small, unresolvable defects creating background at low temperatures, or a combination of the two. As the dose increased to 3 dpa, a more significant difference was observed in the irradiated microstructures over the temperature range. At 300 K, the defects remained as black dot damage. In contrast, during the irradiation at 673 K, small dislocation loops showing double-arc contrast were observed. At 1073 K, these loops were significantly larger, as shown in the bottom right of Figure 3.21, which shows a series of micrographs from ZrC 0.9 irradiated to 3 dpa for each temperature in the range. In every irradiation, a pre-existing dislocation was initially used to follow a certain area. This set of images shows that the dislocations have disappeared more, or completely, at higher temperatures, which could indicate interaction with other lattice defects that become more mobile at higher temperature. The microstructure observed at 5 dpa was strongly dependent on the irradiation temperature. At 300 K, the irradiated microstructure continued to be dominated by

74 62 Figure Dark-field micrographs of ZrC 0.9 irradiated to 1 dpa at temperatures ranging from 300 K to 1073 K. Figure Dark-field micrographs of ZrC 0.9 irradiated to 3 dpa at temperatures ranging from 300 K to 1073 K.

75 63 black-dot damage. At 473 K, small dislocation loops showing double-arc contrast were observed. The same was observed for 673 K at 3 dpa, however that was the maximum dose achieved at that temperature. For the irradiation at 873 K, larger loops not showing double-arc contrast were observed. The loop size continued to increase as the irradiation temperature was increased to 1073 K. At the highest temperatures, a complex dislocation structure has formed in the material. Examples of the final defect structure over the range of temperatures are shown in Figure Figure Bright-field micrographs of ZrC 0.9 irradiated to 5 dpa (except where noted) at temperatures ranging from 300 K to 1073 K. 3.4 Habit Planes of Dislocation Loops at 1073K A series of images taken in a thinner region of ZrC 0.9 after irradiation at 1073 K were particularly useful in examining the nature of the dislocation loops that formed. In this case, the image was taken near the 220 zone axis, as shown in Figure The

76 64 diffraction pattern was indexed to maintain consistency in the labeling of crystallographic directions. The micrograph, presented in Figure 3.24, was obtained at 2 dpa and shows a number of larger loops. Among the loops are several that are seen edge-on, which allow for the determination of the loop habit plane. These edge-on loops were found to lie in {110} and {111} type planes. Knowing that loops exist on these families of planes, and assuming that the dislocation loops are circular, the expected projection could be determined and compared with the image to find the habit plane of additional loops. Both habit planes are listed when they have the same projected shape. These habit planes are consistent with the observation of Frank loops made by Yang, et al. [31] and Frank loops and prismatic loops by Snead, et al. [18]. Figure Figure Diffraction pattern showing the zone axis near the imaging condition for 3.5 Dependence of Irradiated Microstructure on Sample Thickness It was observed at elevated temperatures that the irradiated microstructure depended on the thickness of the sample. This phenomenon can be seen near the edge of

77 65 Figure Dark-field micrograph of ZrC 0.9 irradiated to 2 dpa at 1073 K. Possible dislocation loop habit planes are indicated.

78 66 the sample, as shown in Figure In this image, the thinnest region is at the edge of the sample in the lower-right, and the sample thickness increases away from the edge toward the upper-left. Near the sample edge, there is a region devoid of defects. As the thickness increases, first black-dot damage and then larger, resolvable dislocation loops are present. In the thickest area, the same complex, tangled dislocation structure is observed as in the area followed in Figures 3.13 and Figure A bright-field micrograph of ZrC 0.9 irradiated to 4 dpa at 1073 K taken near the sample edge showing the variation in microstructure with sample thickness. The effect of sample thickness on irradiated microstructure was also well observed down to an irradiation temperature of 873 K. The sample of ZrC 0.8 that was irradiated at 873 K had a particularly large thin area, and a series of micrographs were taken and spliced to follow the evolution of radiation damage with thickness at the terminal dose of 5 dpa. Figure 3.26 is a low magnification image showing the area in which this was performed. The average defect diameter and largest defect diameter were determined as a function of distance away from the edge. This data shows that the average defect

79 67 diameter changes little over this area and increases by approximately 2 nm. The maximum defect diameter, however, increases from about 10 nm to 20 nm, or a four-fold increase in area. It is possible that the foil surface acts as a strong, inexhaustible sink for point defects that limits the formation and growth of defect clusters and dislocation loops near the sample surface. This would result in a defect denuded zone near the sample edge where the distance from a surface is always small. Defects would then appear and be able to grow further away from the sample edge where the foil is thicker such that defects can develop further from the surface. This mechanism would then produce a cross-sectional damage profile as shown in Figure In this case, as you move further from the edge in an image, larger and larger defects would be superimposed over the previous microstructure. The maximum defect size would then increase more quickly than the average. 3.6 Indexing Diffraction Rings/Spots Formed During Irradiation The formation of additional diffraction rings or diffraction spots are observed during many of the irradiations. In fact, the only conditions in which this phenomenon is not observed is at 300 K and 473 K for both stoichiometries. In order to better understand the source of this new diffraction, the ring patterns are indexed. This is often a difficult task due to low intensity; however, by integrating the diffraction pattern in a wedge shape to exclude the ZrC matrix diffracted spots, a 1-D diffraction pattern is created and easily indexed. An example of this process is shown in Figure 3.28.

80 68 Figure Top: A bright-field micrograph of ZrC 0.8 irradiated to 5 dpa at 873 K taken near the sample edge. The marked region shows the analyzed area. Bottom: Quantitative analysis of the average (left) and maximum (right) defect diameter as a function of an arbitrary distance from the edge of the sample.

81 69 Figure A cross-section schematic showing the proposed defect profile due to the action of the sample surface as a point defect sink. Figure Left: A diffraction pattern shows the diffraction rings that develop in ZrC 0.9 irradiated to 5 dpa at 873 K. Right: The diffracted intensity is integrated over a wedge from the DP and the expected FCC lattice peaks with a 0 = 5.06 Å are indicated.

82 70 It is found that the diffraction rings and spots can be attributed to a face-centered cubic (FCC) phase with a lattice parameter of 5.06 Å, or 8% larger than the ZrC matrix. These observations are consistent with that of Gan, et al. [30] who, during an in situ ion irradiation at the IVEM, also found the rings attributed to an FCC phase 8% larger than ZrC. Gosset, et al. [32], however, found diffraction rings develop during bulk irradiation at room temperature but attributed the rings to tetragonal zirconia. Through tilting and dark-field TEM, they found the phase producing the rings to be located on the surfaces pre-thinned before irradiation. It is reasonable to postulate that the surface of the sample oxidizes to form zirconia during irradiation, especially at elevated temperatures. If it was zirconia that had formed, it would not be possible to distinguish between the two phases, tetragonal and cubic, with the data available from this research. The diffraction pattern resolution is too low for this purpose. However, one oddity was that the diffraction rings were also observed at 20 K and 50 K, but not at 300 K or 473 K. ZrC should, typically, be less likely to oxidize as the temperature decreases, which makes the observation of this phase at cryogenic temperatures peculiar. A possible explanation is that oxygen-rich molecules, such as water, adhere to the surface of the sample at cryogenic temperatures and, through a radiation induced process, form zirconia at the interface. To summarize, the diffraction rings or spots formed during irradiation could be attributed to tetragonal or cubic zirconia or another similar-structured phase.

83 Effect of Stoichiometry on Damage Accumulation The effects of stoichiometry were studied by comparison of the irradiated microstructures formed in ZrC 0.8 and ZrC 0.9. The ZrC x phase is stable down to approximately ZrC 0.5, or 33% carbon [10]. This would suggest that the difference in stoichiometry is accommodated by vacancies on the carbon sub-lattice, 20% for ZrC 0.8 and 10% for ZrC 0.9. Figure 3.29 shows a direct comparison using dark-field images taken for both stoichiometries after irradiation to 5 dpa at 473 K and 1073 K. At 473 K, both show a high density of black-dot damage, and at 1073 K, both show a coarse microstructure of tangled dislocations or overlapping loop projections. Over the entire range of temperatures, little difference is observed in the irradiated microstructures, which suggests that the carbon sub-lattice does not have a strong influence on the development of irradiation damage in ZrC. 3.8 Thermal Annealing of ZrC 0.9 Irradiated to 1 DPA at 300 K The irradiation experiments showed a trend of increasing defect size with increasing irradiation temperature. At 1073 K, the irradiated microstructure initially developed as black dot damage and then coarsened into larger dislocation loops with increasing dose. From these observations, it was concluded that both increasing irradiation temperature and increasing dose at high temperature promotes coarsening of the microstructure and leads to larger defects.

84 Figure Dark-field micrographs of ZrC 0.8 (left) and ZrC 0.9 (right) irradiated to 5 dpa at 473 K (top) and 1073 K (bottom) showing similar microstructures. 72

85 73 In an effort to distinguish between these two effects, an experiment was performed in which a sample was irradiated to 1 dpa at room temperature (300 K) and then progressively annealed at 473 K, 673 K, 873 K, and 1073 K for 30 minutes at each temperature. Through this procedure, it would be possible to look for possible activation of defect mobility by observing the microstructure in real time. Micrographs were recorded after annealing at each temperature, and the resultant microstructures could be compared. Finally, the microstructure developed after annealing at 1073 K could be compared to that after irradiating at 1073 K. The microstructure was continuously observed throughout the experiment. There was no visible evidence of defect mobility at any point in the experiment. Changes in contrast of some defects were observed, but this could be attributed to small changes in the deviation parameter (imaging condition). The series of micrographs depicting the anneal progression is shown in Figure Any changes in the microstructure appeared to be minimal up to 873 K. However, after annealing at 1073 K, there did appear to be some coarsening of the microstructure. The contrast of the defects increased and there was a decrease in the background intensity. However, no motion of the visible defects was observed. The microstructure after annealing at 1073 K is compared to that resulting from the irradiation to 1 dpa at 1073 K in Figure The microstructure from irradiation at temperature was coarser than the annealed microstructure. There were more large defects that resulted from the irradiation at temperature and there are clearly defined dislocation loops; some examples are indicated by the arrows. The majority of defects in the annealed sample remained as black dot damage.

86 74 Figure Dark-field TEM images showing the progression of the microstructure through the annealing steps at 473 K, 673K, 873 K and 1073 K.

87 75 Figure Dark-field TEM images comparing the microstructures obtained by annealing at 1073 K after irradiation to 1 dpa at 300 K (top) to that obtained by irradiating to 1 dpa at 1073 K (bottom).