Radiation-induced Dimensional Changes Research Articles

Surface modification of PAN based carbon fibers by high fluence N+ and C+ ion irradiation

Surface structure modification of PAN based carbon fibers under 30 keV C+, N+, N2+ and Ar + ions irradiation have been experimentally studied. It has been found that the corrugation effect of the fiber surface, which is typical for high-fluence irradiation with noble gases and nitrogen ions, does not manifest itself under irradiation with carbon ions. The dependences of the formation of surface patterns on depth distributions ν(x) of the number of displacements per atom (dpa) are considered. It has been shown that during corrugation the ν decreases monotonically with a maximum at the surface. The cause of corrugation is twinning of fiber-shell crystallites, which is caused by depth-dependent radiation-induced dimensional changes and relaxation of resulting mechanical stresses. The reason for the absence of surface corrugation under C+ irradiation can be the formation of a textureless modified layer.

Surface patterns formation by ion irradiation of PAN based сarbon fibers

Surface patterns on PAN based сarbon fibers under 3 and 30 keV He+ and 30 keV Ar+ ions irradiation were experimentally studied. It has been found that the depth distributions ν(х) of the number of displacements per atom (dpa) have a strong effect on the surface patterns on PAN-based carbon fibers. For not steady state profiles ν(х), a ridge-like structure oriented along the fibers is formed on the surface. At sufficiently high irradiation fluences with a steady-state profile ν(х), a corrugated structure is formed with corrugations transverse to the fiber axis. The corrugation effect is explained by twinning during plastic deformation of fiber shell crystallites caused by radiation-induced dimensional changes and mechanical stresses.

The regularities of surface corrugation of polyacrylonitrile based carbon fibers under high-fluence ion irradiation

The regularities of surface corrugation of polyacrylonitrile based carbon fibers reinforcing composite KUP-VM has been studied experimentally after high-fluence 30 keV Ne+, 20 and 30 keV Ar+, 30 keV N2+ ion irradiation at target temperatures ranging from 100 to 600 °C. It was found that periodic corrugated structure parameters, such as the angles of the inclination β of the corrugations and their fraction on the fiber surface have strong temperature dependences. The periodicity of corrugations for all used ions slightly decreases with increasing temperature and is of the order of hundreds of nanometers. For neon and argon ions the corrugation inclination angles and the fraction in contrary of nitrogen ions show minima at the temperatures near 400 °C which corresponds to the minimum of radiation-induced dimensional changes in carbon materials. The depth of the minima at 400 °C depends on the competition of the processes of ion-induced plastic deformation of carbon materials and ion bombardment sputtering. For nitrogen ions the inclination angles increase with temperature so that at temperature ∼270 °C they reach maximum values ∼40°, staying at this level at subsequent temperature elevation up to 600 °C. The qualitative difference in the temperature dependences of fiber corrugation for noble gas ions and nitrogen ones is connected to nitrogen chemical activity and the creation of CN bonds under nitrogen ion implantation.

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.

On the nature of cracks and voids in nuclear graphite

Microcracks in neutron-irradiated nuclear grade graphite have been examined in detail for the first time using a combination of transmission electron microscopy (TEM), electron energy loss spectroscopy (EELS), energy dispersive X-ray (EDX), and energy filtered TEM (EFTEM). Filler particles from both unirradiated Pile Grade A (PGA) and three irradiated British Experimental Pile ‘O’ (BEPO) graphite specimens were investigated with received doses ranging from 0.4 to 1.44 displacements per atom (dpa) and an irradiation temperature of between 20 and 120 °C. We suggest that the concentration and potentially the size of microcracks increase with increasing neutron irradiation and show that disordered carbon material is present in a range of microcracks (of varying size and shape) in all specimens including unirradiated material. EFTEM and EELS data showed that these cracks contained carbon material of lower density and graphitic character than that of the surrounding bulk graphite. The presence of partially filled microcracks has potentially significant implications for the development of microstructural models for the prediction of radiation-induced dimensional and property changes in nuclear graphite.

Carbon–Carbon Composite Materials

Data are summarized on the performance characteristics and properties of Russian-produced and foreign carbon–carbon composite materials for various engineering applications. The effect of neutron irradiation on their macroscopic properties and structure is examined. The relationships between the radiation-induced dimensional changes and properties of the composites are established, which can be used in assessing the engineering performance and optimizing the fabrication of carbon–carbon composites.

Comprehensive structural analysis of the HCPB DEMO blanket under thermal, mechanical, electromagnetic and radiation induced loads

For the helium-cooled pebble bed (HCPB) blanket, which is one of the two reference concepts studied within the European Demo Development Program, a comprehensive finite element (FEM) structural analysis has been performed. The analysis refers to the steady-state operating conditions of an outboard blanket segment. On the basis of a three-dimensional model of radial–toroidal sections of the segment box, thermal stresses caused by the temperature gradients in the blanket structure have been calculated. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions as well as an accidental over-pressurization of the blanket box have been accounted for. The stresses caused by a central plasma major disruption from an initial current of 20 MA to zero in 20 ms have been also taken into account. Radiation-induced dimensional changes of breeder and multiplier material caused by both helium production and neutron damage, have also been evaluated and discussed. All the above loads have been combined as input for a FEM stress analysis and the resulting stress distribution has been evaluated according to the American Society of Mechanical Engineers (ASME) norms.

The reaction rate theory of radiation effects

In materials irradiated with energetic neutrons or charged particles, point defects, clusters, and extended-defect microstructures are produced. Because of the possibility of complete atom-by-atom rearrangement of the material under irradiation, significant changes may result in physical, mechanical, and electrical properties. The theoretical and computational approaches for describing these changes begin with the atomic displacement reactions and follow numerous reaction pathways to an evolved microstructure and the properties dictated by it. Rate theory based on defect reactions has been developed to understand these processes. It is the only approach that spans the large time, spatial, and energy ranges from defect production to changes in macroscopic properties. It has been most thoroughly applied to the phenomena of radiation-induced dimensional changes, particularly swelling and creep. This emphasis has derived both from the importance of these phenomena in technological applications and from their scientific challenge.

Theory and experimental background on dimensional changes in irradiated alloys

When a material is irradiated with energetic neutrons or charged particles, a complex sequence of reactions takes place. Structure, composition, and properties are altered over an extremely wide scale, spanning atomic defects, meso-scale microstructures and macroscopic properties. Particularly interesting are radiation-induced dimensional changes. Such changes, which can occur in engineered components of fission and fusion reactors on a scale of meters, are driven by lattice defects at the subnanometer level. Because of their technological importance and their high scientific challenge, the dimensional changes termed radiation-induced swelling and creep have elicited sustained intensive research by basic and applied materials scientists for many years. The present paper is intended as a brief tutorial on salient features of this work. The presentation is divided into three parts. Background is first sketched emphasizing experimentally observed features and applications. Next, the theoretical framework and specific models that have been developed to understand radiation-induced swelling and creep in isotropic materials are described. Lastly, selected experiments designed and/or interpreted in terms of theory are highlighted to illustrate the current state of understanding of the physical bases of these phenomena.

Neutron irradiation-induced dimensional changes in a cesium-highly oriented pyrolytic graphite intercalation compound

Direct observations of Cs-intercalated, highly oriented pyrolytic graphite (HOPG) exposed to neutron irradiation show that the presence of Cs dramatically enhances the radiation-induced dimensional changes. A qualitative explanation of this behavior employs an island domain or “pleated layer” model of the graphite intercalation compound (GIC) structure. The inferences from this model are consistent with earlier data for neutron-induced dimensional changes in oriented pure graphite as a function of crystallite size.

Capacitance dilatometer for i n s i t u measurement of radiation-induced dimensional changes of solids

We describe a new capacitance dilatometer for measuring radiation-induced volume change of solids having high stability and capability of measuring a fractional length change of 10−6.

Radiation-induced dimensional changes in zirconium

Experimental studies of radiation-induced dimensional changes in zirconium have been undertaken in the RA1 reactor of the CNEA in Buenos Aires. Both undoped and 235U-doped strip specimens were prepared with preferred orientation to obtain growth in the rolling direction. Bimetal strips were prepared by recombining doped and undoped samples. During bombardment, in the mixed spectrum of the RA1, fission in doped specimens increased the damage rate by nearly two orders of magnitude compared with the undoped. The greater dimensional changes occurring in the doped specimen produce a bow in the bimetal. Measuring the bowing amplitude enables the calculation of relative dimensional changes. Results obtained from bombardments at 365 K in a high temperature loop and at 77 K in a cryostat will be presented and discussed in detail. It is of particular significance to emphasize that at 77 K the dimensional changes were negligible, suggesting that defect migration plays a significant role in the induced dimensional changes.

Radiation-induced dimensional changes and creep in carbonaceous materials

At high fast neutron exposures, the dependences of the irradiation-induced dimensional changes on the structure of turbostratic carbons depart from those established at lower exposures. After an initial densification which is first order with respect to the density defect, the densification ceases and the material expands. Concomitantly, the rate at which the material changes shape increases. It is thought that the acceleration of the rate of dimensional change is a result of the autocatalytic influence of radiation-generated defects and irradiation-induced increases in anisotropy. The new porosity responsible for the expansion at high doses is thought to be associated with large intercrystalline strains that are relaxed by radiation creep. Measurements of the macroscopic creep strains indicate that strain rates of at least about 2, 3 and 6% per 10 21 n/ cm 2 ( E >> 0.18 MeV) can be accommodated at 600, 900 and 1200 °C, respectively, and total strains of about 10% in tension are possible. These measurements, together with the dimensional change behavior of isotropic carbon coatings for nuclear fuel particles, show that for survival to high neutron exposures, the density must be chosen so that coatings will not fail early in life due to densification and so that intolerable anisotropic behavior will not be encountered late in life. The data suggest that, for the designs most commonly used, the optimum densities for isotropic carbons lie in the range 1.5 to 1.8 g/cm 3. Comparison of the data for turbostratic carbons and data for graphitic materials shows that at approximately 1200 °C, the dimensional change rate can vary by two orders of magnitude with crystallinity. The influence is considerably less at 600 °C.

Dimensional changes in graphite irradiated between 500° and 1100°C

Two previously proposed quantitative models for the radiation-induced dimensional changes in polycrystalline graphite treated the material as containing a single crystalline phase. One model was based on thermodynamic relationships and the other was derived by averaging the contributions of the individual crystallites. A third model emphasized the effects of small quantities of a second, less crystalline phase, but this model has not been expressed analytically. In the present work, measurements of dimensional changes in radial, circumferential, and longitudinal samples from an extruded graphite tube irradiated to exposures between 1.5 and 2.4 × 10 21 nvt (E>0.18 MeV) at temperatures between 540° and 1060°C were compared with the relationships predicted by the two quantitative models. It was found that the data agreed best with the single-phase thermodynamic model.

Radiation-induced dimensional changes in pyrolytic carbons deposited in a fluidized bed

A range of pyrolytic carbon structures representative of those used as coatings for nuclear fuel particles was irradiated at temperatures between 540° and 1060°C to fast neutron exposures of 1.5 to 2.4 × 10 81 nvt ( E > 0.18 MeV). Changes in linear dimensions, density, and crystallite size were measured. The data show that irradiation-induced dimensional changes in pyrolytic carbons depend sensitively on their structure. At ~1000°C the dimensional changes depend primarily on the degree of preferred orientation with a secondary dependence on the density of the carbon. The changes calculated for the crystallites are independent of the crystallite height, L c , in the range 30 to 145 Å but the changes are less for larger L c values. At ~600°C the dimensional changes depend on the L c , the density, and the degree of preferred orientation. At this temperature the calculated crystallite changes are very sensitive to variations in L c in the range 30 to 200 Å. At all temperatures the irradiation causes the carbons to densify. The irradiation tends to increase the L c value of carbons with initially small L c values and decreases those of carbons with initially high L c values. The observed irradiation-induced changes are discussed in terms of current theories of irradiation damage in carbons and graphite.

Notes on the two-phase model for radiation-induced dimensional changes in graphite

The historical development and modifications to the “single-phase model” of the graphite structure, used to interpret and predict behavior during neutron irradiation, are outlined. In the singlephase model all parts of the artificial graphite aggregate are assumed to behave as if they were part of a graphite crystal lattice. It is shown that changes occurring in the non-graphitic carbon phase, relief of internal stresses induced during cool-down from graphitization temperature, and chances in the crack structure of the graphite aggregate must also be considered in interpretation of dimensional changes produced during high-temperature irradiations.

The “size effect” of radiation-induced dimensional changes of graphite

The “size effect” of radiation-induced dimensional changes of graphite

Radiation-Induced Dimensional Changes in Large Graphite Bars

The radiation-induced contraction of large graphite bars irradiated at high temperatures is approximately twice that of the small cylindrical samples that are conventionally studied. Although the r...

Effect of low compressive stresses on radiation-induced dimensional changes in graphite

When graphite is employed as a reactor moderator, the variations of temperature and neutron flux within the moderator bars cause internal stresses to be generated in the graphite. The effect of these stresses on the dimensional changes of the graphite are of immediate interest in the field of reactor design. Samples of five nuclear-grade graphites have been irradiated under constant compressive stresses of 150 and 300 psi. Length changes of the stressed samples are compared with length changes of unstressed control samples, irradiated in the same assembly. Results obtained at 625°C and for exposures up to 6500 MWd/At in a Hanford irradiation facility are presented. Additional data on one nuclear-grade graphite, irradiated under 100 psi constant tensile stress, are included.

32. Effect of low compressive stresses on radiation-induced dimensional changes in graphite

32. Effect of low compressive stresses on radiation-induced dimensional changes in graphite