Irradiation-induced Shrinkage Research Articles

Comprehensive characterization of the irradiation effects of glassy carbon

Carbon materials have become increasingly diverse, finding applications in high-temperature and high-radiation environments. Glassy carbon, an allotrope known for its exceptional chemical inertness and desirable mechanical properties. However, understanding neutron irradiation effects in glassy carbon has proven challenging, primarily because of its unique nanopore structure. This study presents a highly detailed microstructural characterization investigation of neutron-induced changes in glassy carbon, revealing how changes in nanopore structure and crystallinity impact the irradiation-induced shrinkage. Aberration-corrected scanning transmission electron microscopy (STEM) reveals pore closure that leads to material densification in the irradiated samples. Dimensional analysis combined with comparison to historical data suggest significant length shrinkage to occur. Neutron and in situ electron irradiation experiments suggest that glassy carbon transforms into so-called carbon onions, supporting the concept of shrinkage saturation. Investigating irradiation temperature effects using STEM, electron energy loss spectroscopy, x-ray diffraction, and Raman spectroscopy revealed partial amorphization at 210 °C–230 °C and preserved order in glassy carbon at 860 °C, coinciding with pore closure. Thermal property measurements were also conducted to assess the effects of densification and other changes in the atomic structure of glassy carbon. The results of this study have broad implications in the deployment of glassy carbon to nuclear environments, based around the observed changes in the thermal properties, and demonstrates the operational window for the onset of densification.

Phase field modeling of irradiation-induced shrinkage fracture in TRISO fuel particle

With the features in retaining fission products and maintaining structural stability, TRISO fuel particle has been widely implemented and studied in the development of advanced nuclear fuel element. During service, however, the key silicon carbide (SiC) layer of TRISO may fail due to irradiation-induced shrinkage fracture of inner pyrolytic carbon (IPyC) layer. Hence, it has great significance to analyze the fracture characteristics of TRISO for its application in engineering. This work presents a phase field model of brittle fracture for the shrinkage cracking, which is capable of simulating the propagation of IPyC crack and interfacial crack between IPyC and SiC naturally. At first, the physical model of TRISO and phase field model have been verified respectively. Subsequently, the crack evolution has been investigated under different parameters, including the Bacon Anisotropy Factor (BAF) of IPyC, fuel temperature and density of IPyC. Simulation results reveal that increasing the BAF or decreasing the density and temperature, the fracture of IPyC will be accelerated and the crack deflection may also occur subsequently. In contrast, the fracture of IPyC can be put off and relieve the stress concentration in SiC. The combination effect of shrinkage and stress relaxation from creep ultimately determines the crack evolution, which in turn influences the stress distribution within SiC.

Irradiation-induced shrinkage of highly crystalline SiC fibers

Composites consisting of a SiC matrix prepared by chemical vapor infiltration that was reinforced by Tyranno™ SA3rd (TSA3) or Hi-Nicalon™ Type S (HNLS) SiC fibers were irradiated with 5.1MeV Si ions to 100dpa at 300°C. It was confirmed that the irradiated microstructure of TSA3 was stable up to 100dpa. In contrast, the results of surface profilometry and cross-sectional transmission electron microscopy (TEM) showed that HNLS fibers underwent 0.8% shrinkage along the axis and 0.7% radial shrinkage. High-resolution TEM images of the highly damaged regions in the HNLS material revealed the complete loss of carbon ribbons initially distributed at the SiC grain boundaries. Ion-beam-induced diffusion of carbon into the SiC grains was indicated from the observations of the interdiffusion layer formed at the fiber/pyrocarbon interface. The π∗ peak of the carbon K-edge spectrum was found in the electron-energy-loss spectrum of the HNLS SiC grains that were irradiated above 60dpa; this peak was not observed in unirradiated SiC fibers, further demonstrating that carbon sp2 bonding was only detected in HNLS SiC fibers. According to the changes in the bonding and the volume relaxation related to the carbon defects in SiC, an increase in the population of antisite CSi was likely the key underlying mechanism of the irradiation-induced shrinkage of the HNLS fibers.

Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

Irradiation-induced shrinkage and expansion mechanisms ofSiO2 circle membrane nanopores

20 nm diameter SiO2 nanopore arrays on gradient-thickness membranes were formed by a focused electron beamwith in situ transmission electron microscopy (TEM). Nanopore shrinkage was seen innanopores on thicker membranes, with the rate of diameter change remaining constantduring the shrinkage process. In contrast, pore expansion was observed in thinnermembranes, with the expansion rate being constant at the initial stage but with a slightincrease at the later stage. The geometry model of shrinkage and expansion of thenanopores in relation to the electron irradiation time was investigated by utilizing the TEMtilting method.

An evaluation of the effects of SiC layer thinning on failure of TRISO-coated fuel particles

The fundamental design for a gas-cooled reactor relies on the behavior of the coated particle fuel. The coating layers surrounding the fuel kernels in these spherical particles, consisting of pyrolytic carbon and silicon carbide layers, act as a pressure vessel that retains fission products. Many more fuel particles have failed in US irradiations than would be expected when only one-dimensional pressure vessel failures are considered. Several multi-dimensional failure mechanisms that may have contributed to these failures have been previously studied, such as (1) irradiation-induced shrinkage cracks in the inner pyrocarbon (IPyC) layer, (2) partial debonding between the IPyC and SiC layers, and (3) deviations from a perfectly spherical shape. A further phenomenon that could lead to particle failures is thinning of the SiC layer caused by either thermal decomposition or interaction with fission products. Results of a study of the effects of SiC thinning and criteria for evaluating this behavior in a fuel performance code are presented.

Consideration of the effects of partial debonding of the IPyC and particle asphericity on TRISO-coated fuel behavior

The fundamental design for a gas-cooled reactor relies on the behavior of the coated particle fuel. The coating layers surrounding the fuel kernels in these spherical particles, consisting of pyrolytic carbon and silicon carbide layers, act as a pressure vessel that retains fission product gases. Many more fuel particles have failed in US irradiations than would be expected when only one-dimensional pressure vessel failures are considered. Post-irradiation examinations indicate that multi-dimensional effects may have contributed to these failures, such as (1) irradiation-induced shrinkage cracks in the inner pyrocarbon (IPyC) layer, (2) partial debonding between the IPyC and SiC layers, and (3) deviations from a perfectly spherical shape. An approach that was used previously to evaluate the effects of irradiation-induced shrinkage cracks is used herein to assess the effects of partial debonding and asphericity. Results of this investigation serve to identify circumstances where these mechanisms may contribute to particle failures.

Statistical approach and benchmarking for modeling of multi-dimensional behavior in TRISO-coated fuel particles

The fundamental design for a gas-cooled reactor relies on the behavior of the coated particle fuel. The coating layers, termed the TRISO coating, act as a mini-pressure vessel that retains fission products. Results of US irradiation experiments show that many more fuel particles have failed than can be attributed to one-dimensional pressure vessel failures alone. Post-irradiation examinations indicate that multi-dimensional effects, such as the presence of irradiation-induced shrinkage cracks in the inner pyrolytic carbon layer, contribute to these failures. To address these effects, the methods of prior one-dimensional models are expanded to capture the stress intensification associated with multi-dimensional behavior. An approximation of the stress levels enables the treatment of statistical variations in numerous design parameters and Monte Carlo sampling over a large number of particles. The approach is shown to make reasonable predictions when used to calculate failure probabilities for irradiation experiments of the New Production – Modular High Temperature Gas Cooled Reactor Program.

Dimensional stability and tensile strength of irradiated Nicalon-CG and Hi-Nicalon fibers

Nicalon-CG and Hi-Nicalon fibers were characterized by measuring their density and tensile strength in the unirradiated, thermal annealed, and irradiated conditions. The results indicate the fibers that perform best after irradiation to 43 dpa SiC at 1000°C are those that approach stoichiometric and crystalline SiC. Hi-Nicalon fiber exhibited less than 1% densification, accompanied by a slight increase in tensile strength after irradiation. Nicalon-CG, in contrast, was significantly weakened in the annealed and irradiated conditions. In addition, Nicalon-CG exhibited substantial irradiation-induced shrinkage. Loss of fiber tensile strength after irradiation is shown to reduce the flexural strength of irradiated composites while fiber shrinkage, and resultant debonding from the matrix, are linked to a reduced composite elastic modulus.

Dimensional Stability and Strength of Neutron-Irradiated SiC-Based Fibers

A variety of SiC-based fibers were characterized by measuring their length, density, and tensile strength in the unirradiated, thermal annealed, and irradiated conditions. The irradiation was conducted in the EBR-II to a dose of 43 dpa-SiC (185 EFPD) at a nominal irradiation temperature of 1000{degree}C. The annealed specimens were held at 1010{degree}C for 165 days to approximately duplicate the thermal exposure of the irradiated specimens. In general, the results of this study indicate the fibers that perform best in an irradiation environment are those that approach stoichiometric and crystalline SiC. Hi-Nicalon exhibited negligible densification, accompanied by an increase in tensile strength after irradiation. Nicalon CG possessed a higher tensile strength than Hi-Nicalon in the unirradiated condition, but was significantly weakened in the annealed and irradiated conditions. In addition, Nicalon CG exhibited unacceptable irradiation-induced shrinkage. While the irradiation stability of Hi-Nicalon was promising, other fibers with compositions closer to stoichiometric SiC may perform even better. This potential was suggested by the MER99 fiber, which displayed excellent dimensional stability. The principal drawback for the fully crystalline and stoichiometric fibers such as MER99 and Crystalline SiC is their low strength and flexibility caused by high flaw concentrations. 6 refs., 11 figs., 5 tabs.

Size effect on the irradiation performance of coated fuel particles

Outer coatings that were as near alike as possible were applied to two different sizes of inert TRISO particles that were larger than those commonly used to fuel HTGR reactors and these particles were then irradiated in a test reactor to observe the influence of particle size on outer coating failures that resulted from irradiation-induced shrinkage of coatings onto the more stable SiC substrates over which they were applied. Outer coatings of plain pyrocarbon and of Si-alloyed pyrocarbon were used to make up two test pairs of particles with diameters of about 1050 and 1300 μm. For a fast-neutron fluence of 5.5 × 10 25 n/m 2 ( E > 29 fJ) at an irradiation temperature of 1125K, failure was about twice as high in the larger 1300 μm particle of each test pair as in the smaller 1050 μm particle (16 vs 8%), with each of the coating types having roughly the same behavior. This observed size effect is somewhat greater than predicted by volume-dependent Weibull theory, which estimates failure of the larger size particles at 13% when the smaller size particles fail at a rate of 8%. However, experimental uncertainties are sufficient to account for the difference between observed size effects and those predicted by Weibull theory. The conclusion that the irradiation performance of coated particles is size dependent is reinforced by the fact that failure for regular 850 μm fueled particles in the same irradiation test was essentially zero.

Behavior of pyrolytic carbons irradiated under mechanical restraint

The behavior of isotropic pyrolytic carbons irradiated free from stress is compared with that of the same carbons when the irradiation-induced shrinkage is prevented by an unyielding substrate. Failure of the restrained specimens cannot be predicted from the irradiation-induced dimensional changes of the unrestrained specimens, and fracture strengths do not appear to decrease during irradiation in either a restrained or an unrestrained state. However, restraint and the resulting irradiation-induced creep produce an increase in the preferred orientation of the carbon cystallites. This increase in preferred orientation results in increased irradiation-induced dimensional change rates which can explain the discrepancies in the irradiation behavior of the restrained and unrestrained carbons.