Irradiation growth (due to anisotropic nucleation and growth of dislocation loops on different habit planes) can be of significant practical concern at intermediate temperatures in anisotropic materials such as Zr alloys, Be, BeO, Al2O3, uranium, and graphite.40’126’259-261 Anisotropic growth in individual grains in polycrys­talline materials can produce large grain boundary stresses, leading to loss of strength and grain bound­ary fracture in some materials. Figure 27 shows the large anisotropy in measured lattice parameter change in the basal and prism planes for BeO irradiated near room temperature.262 For neutron fluences above 2 x 1020 cm-2 (~0.2 dpa) with a c-axis expansion >0.5% and an a-axis expansion near 0.1%, a rapid decrease in flexural strength was observed.262,263 In materials with highly textured grains, unacceptable anisotropic growth at the mac­roscopic level can occur. One engineering solution is to use processing techniques to produce randomly aligned, small grain-sized materials.

Irradiation creep occurs in the presence ofapplied stress, due to biased absorption of point defects at cavities and along specific dislocation orientations relative to the applied stress.264 Irradiation creep produces dimensional expansion that acts in addition to normal thermal creep mechanisms and is most

prominent at temperatures from recovery Stage III up to temperatures where thermal creep deformation becomes rapid (typically above 0.5 TM). The magni­tude ofsteady-state irradiation creep is proportional to the applied stress level and dose, and consists of a creep compliance term and a void swelling term. The magnitude of typical irradiation creep compli­ance coefficients260,265,266 for fcc and bcc metals

is 0.5-1 x 10-12 Pa-1 dpa-1. The irradiation creep compliance for ferritic/martensitic steels appears to be about one-half of that for austenitic steels.109 Accelerated irradiation creep due to differential absorption of point defects at low temperatures (e. g. below recovery Stage V) or at low doses can produce creep deformation rates that are up to 10-100 times larger than the steady-state irradiation

267,268

creep rates.

Volumetric swelling from void formation occurs at temperatures above recovery Stage V in fcc and HCP materials (and above Stage III for bcc materi­als), and typically exhibits a linear increase with dose after an initial transient regime. As summarized in Figure 28 the dose-dependent swelling in fast fission reactor-irradiated austenitic stainless steel progresses

at a swelling rate of ^1%/dpa without evidence for saturation up to swelling levels approaching 100%.1 Similar high swelling levels without evidence of sat­uration have been observed in pure copper108 and some simple bcc alloys.131 Volumetric swelling levels in structural materials in excess of ^5% are difficult to accommodate by engineering design,269 and addi­tional embrittlement mechanisms may appear in aus­tenitic stainless steel for swelling levels above 10% including void channeling and loss of ductility.270,271 Therefore, there is strong motivation to design struc­tural materials that are resistant to void swelling by introducing a high matrix density of point defect sinks or other techniques. In general, the amount of void swelling is lower in bcc materials compared to

fcc materials.50,92,109 For example, the observed void

swelling in many ferritic/martensitic steels is <2% after fission neutron damage levels of 50 dpa or higher, whereas the void swelling in simple austenitic stainless steels may be 30% or higher.109 The superior swelling resistance in ferritic/martensitic steels is largely due to a higher transient dose before onset of steady-state swelling, along with a lower steady-state swelling rate. For many HCP materials, the amount of void swelling is relatively small compared to fcc materials due to anisotropic point defect migration that tends to pro­mote defect recombination.128 However, the potential for anisotropic swelling associated with cavity formation in HCP materials may induce large stresses and poten­tial cracking at grain boundaries.263,272,273 Figure 29 shows an example of aligned cavity formation and grain boundary separation in Al2O3 following fast fis­sion reactor irradiation.272