E.P. Simonen

Radiation-induced material changes and susceptibility to intergranular failure of light-water-reactor core internals

Abstract Current understanding of radiation-induced material changes that occur in light-water-reactor (LWR) core components is critically reviewed and linked to intergranular failure processes. Although the basic science of radiation damage processes in metals is reasonably well established, accurate prediction of microstructures, microchemistries and mechanical property changes in complex stainless alloys during irradiation at LWR temperatures is not possible at present. Mechanistic understanding of these radiation-induced changes in commercial alloys is considered to be of paramount importance for the mitigation of the intergranular environmental cracking that occurs in service. Fundamental research is needed to define defect–solute interactions and microstructural evolution at intermediate temperatures and dose rates pertinent to LWRs where transient effects often dominate behavior. In addition, it is essential that radiation effects on matrix microstructure and microchemistry and grain boundary microchemistry be understood. Finally, a stronger emphasis on accurately quantifying radiation effects on environmental cracking mechanisms and kinetics is needed.

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.

Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 C

Six austenitic stainless steel heats (three heats each of 304SS and 316SS) neutron-irradiated at 275 °C from 0.6 to 13.3 dpa have been carefully characterized by TEM and their hardness measured as a function of dose. The characterization revealed that the microstructure is dominated by a very high density of small Frank loops present in sizes as small as 1 nm and perhaps lower, which could be of both vacancy and interstitial-type. Frank loop density saturated at the lowest doses characterized, whereas the Frank loop size distributions changed with increasing dose from an initially narrow, symmetric shape to a broader, asymmetric shape. Although substantial hardening is caused by the small defects, a simple correlation between hardness changes and density and size of defects does not exist. These results indicate that radiation-induced segregation to the Frank loops could play a role in both defect evolution and hardening response.

Fluence and temperature dependence of swelling in irradiated molybdenum

Abstract Volume changes have been analyzed in molybdenum which has been neutron irradiated to various fluences over the temperature range 50 to 1300°C. This data together with all previously reported data has been compiled into a three-dimensional plot of swelling versus temperature versus fluence. Significant low temperature swelling, 400°C. The nature of the dislocation and void microstructure at high irradiation temperatures are analyzed quantitatively as a function of irradiation temperature and the results are reasonably consistent with a recent model of Brailsford and Bullough. The same model is also consistent with an observed trend towards saturation in the void swelling at high fluences at the low temperature end of the void region.

Effect of helium on void formation in nickel

Abstract This study examines the influence of helium on void formation in self-ion irradiated nickel. Helium was injected either simultaneously with, or prior to, the self-ion bombardment. The void microstructure was characterized as a function of helium deposition rate and the total heavy-ion dose. In particular, at 575°C and 5 × 10 −3 displacements per atom per second the void density is found to be proportional to the helium deposition rate. The dose dependence of swelling is initially dominated by helium driven nucleation. The void density rapidly saturates after which swelling continues with increasing dose only from void growth. We conclude that helium promotes void nucleation in nickel with either helium implantation technique, pre-injection or simultaneous injection. Qualitative differences, however, are recognized.

The influence of oversized solute additions on radiation-induced changes and post-irradiation intergranular stress corrosion cracking behavior in high-purity 316 stainless steels

The influence of oversized solute additions on the radiation-induced microstructure, radiation-induced segregation (RIS) at grain boundaries and post-irradiation intergranular stress corrosion cracking (IGSCC) behavior of model, high-purity 316 stainless alloys, doped with either 0.3 at.% platinum or 0.3 at.% hafnium, and proton-irradiated to 2.5 and 5 dpa at 400 °C was examined. Radiation-induced microstructure was characterized using both bright and dark field imaging techniques in transmission electron microscopy. Platinum addition was found to promote void nucleation and to increase both the loop density and the mean loop diameter relative to the base alloy at 2.5 dpa. Addition of hafnium was effective in reducing swelling at 2.5 and 5 dpa. Hafnium addition also significantly decreased the mean loop diameter relative to the base alloy. Both platinum and hafnium additions also resulted in significant suppression of RIS at grain boundaries at 2.5 dpa. At 5 dpa, the influence of hafnium addition on RIS was still beneficial but much less pronounced. Comparative constant elongation rate tensile tests performed in a simulated boiling water reactor environment at 288 °C demonstrated a beneficial effect of hafnium addition and to a lesser extent platinum addition on the post-irradiation IGSCC behavior of 316 stainless steel alloys. The 316SS alloy doped with platinum exhibited a slightly lower susceptibility to post-irradiation IGSCC than the 316SS base alloy at both 2.5 and 5 dpa. Most spectacularly, the 316SS alloy doped with hafnium was found to be not susceptible to post-irradiation IGSCC at both 2.5 and 5 dpa. The mechanisms by which oversized solute additions impact point defect behavior as well as the links between radiation-induced changes and irradiation-assisted stress corrosion cracking are discussed.

Lattice defect/grain boundary interactions related to IASCC

Abstract Radiation-induced segregation (RIS) of major alloying elements to grain boundaries in austenitic stainless steels has emerged as a critical aspect of irradiation-assisted stress corrosion cracking (IASCC). Discriminating interactions between individual solute species and vacancy and interstitial defects as they migrate to grain boundaries result in redistribution of solute. Measurements of grain boundary Ni enrichment and Cr depletion indicate that RIS of major alloying elements is in reasonable agreement with the inverse-Kirkendall mechanism. The discriminating interactions for inverse-Kirkendal segregation are the relative rates of solute diffusion by vacancy exchange. Mechanistically, the ternary composition path, defined by change in Cr relative to Ni, depends on relative diffusivities. The absolute change in the composition, that is, extent along the composition path, depends on the kinetics of vacancy formation and migration. The composition path approach is used to quantify diffusional characteristics at low temperatures. Lastly, model predictions suggest a significant influence of grain boundary defect characteristics in addition to matrix defect characteristics. These grain-boundary sensitive characteristics may influence IASCC.

Light ion irradiation-induced creep mechanisms in nickel

Abstract Irradiation creep of fusion reactor structural materials is important for considerations of dimensional stability and creep rupture failure predictions. In the present work, the predicted creep response at 200°C of both cyclic and steady-state deuteron irradiated nickel is analyzed and compared with experiment. The cycle characteristics were chosen to be similar to that expected for a tokamak, i.e., 1000 seconds irradiation followed by 100 seconds at temperature without irradiation. The cyclic creep rate is observed to be approximately three times the creep rate with continuous irradiation. The analysis of both the continuous beam and cyclic beam results support the conclusion that the radiation induced creep mechanism is enhanced climb-glide creep and not stress induced preferential absorption (SIPA).

Evidence for excess vacancies at sliding grain boundaries during superplastic deformation

Rapid quenching of Al-Mg alloys during superplastic deformation has revealed the presence of nano-scale cavities along many grain boundaries. They were observed only under deformation conditions where grain boundary sliding was the dominant mechanism. Fine-probe compositional measurements revealed that the cavity surface is enriched in Mg, and in situ heating in the transmission electron microscope demonstrated that they are not stable above 175 C. Kinetic analysis of cavity formation during a quench concludes that the cavities did not exist during deformation but were formed as the sample cooled. It is proposed that these cavities are evidence for a localized excess of vacancies during grain boundary sliding.

Pulsed flux effects on radiation damage

Abstract Pulsed irradiation fluxes can cause alteration in the development of irradiation microstructures when compared to steady irradiation microstructures. Theoretical analysis of irradiation damage development and examination of pulsed ion irradiation microstructures have indicated conditions for expected pulsing effects on fusion reactor materials. In pure metals, high instantaneous damage rates and pulse annealing periods comparable to defect relaxation times can cause significant pulsing effects. In addition, irradiation affected phases in alloys are altered by pulsed irradiation compared to steady irradiation.