Modeling dislocation structure development and creep-swelling coupling in neutron irradiated stainless steel

Anisotropic nucleation and growth of multi-classes of dislocation loops under the combined actions of fast-neutrons and an external applied stress are considered in modeling dislocation structure development in metals and alloys. The stochastic nature of the nucleation kinetics is formulated via the Fokker-Planck equation. The strain derived from the climb of the anisotropic dislocation structure is separable into volumetric and deviatoric components, corresponding respectively to swelling and creep. The creep contribution resulting from the development of the stress-induced dislocation anisotropy is found to be very significant and exhibits a strong correlation with swelling. For stainless steel, our model explains very well the complex deformation behavior observed in a wide variety of in-reactor experiments.     

Fracture behavior of austenitic stainless steels irradiated in PWR

Fracture behavior of cold-worked 316 stainless steels irradiated up to 73 dpa in a pressurized water reactor was investigated by impact testing at −196, 30 and 150 °C, and by conventional tensile and slow tensile testing at 30 and 320 °C. In impact tests, brittle IG mode was dominant at −196 °C at doses higher than 11 dpa accompanying significant decrease in absorbed energy. The mixed IG mode, which was characterized by isolated grain facets in ductile dimples, appeared at 30 and 150 °C whereas the fracture occurred macroscopically in a ductile manner. The sensitivity to IG or mixed IG mode was more pronounced for higher dose and lower test temperature. In uniaxial tensile tests, IG mode at a slow strain rate appeared only at 320 °C whereas mixed IG mode appeared at both 30 and 320 °C at a fast strain rate. A compilation of the results and literature data suggested that IG fracture exists in two different conditions, low-temperature high-strain-rate (LTHR) and high-temperature low-strain-rate (HTLR) conditions. These two conditions for IG fracture likely correspond to two different deformation modes, twining and channeling.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Tomographic atom probe characterization of the microstructure of a cold worked 316 austenitic stainless steel after neutron irradiation

For the first time, chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after 17 years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa (7.81 × 10²⁵ neutrons.m⁻², E > 1 MeV). The samples were analysed with a laser assisted tomographic atom probe. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A high number density (6 × 10²³ m⁻³) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. Mo and P segregations at the interfaces of these clusters were also observed. Finally, Si enriched atmospheres were seen.     

Microstructural evolution of the China Low Activation Martensitic (CLAM) steel irradiated by H and He ion beams

China Low Activation Martensitic (CLAM) steel was irradiated at room temperature with different doses of He⁺ and H⁺ ion beams. TEM indicated that the microstructure of unirradiated CLAM steel consisted of laths, grain boundaries, dislocations and carbides. Electron diffraction patterns revealed that the microstructure of carbides at grain boundaries was primarily dominated by M₂₃C₆ carbide. Vacancy clusters were induced into the matrix after irradiation. TEM-EDX of carbides and matrices of unirradiated and post-irradiated samples were performed to investigate the composition of carbides and the effect of irradiation on the composition of carbides. Carbides from unirradiated and irradiated specimens at grain boundaries were found to be enriched with Cr. For irradiated specimens, concentrations of Cr increased as the irradiation dose was increased. Cr enrichment could lead to precipitation of additional phase.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

A polycrystalline modeling of the mechanical behavior of neutron irradiated zirconium alloys

Zirconium alloys used as fuel cladding tubes in the nuclear industry undergo important changes after neutron irradiation in the microstructure as well as in the mechanical properties. However, the effects of the specific post-irradiation deformation mechanisms on the mechanical behavior are not clearly understood and modeled. Based on experimental results it is discussed that the kinematic strain hardening is increased by the plastic strain localization inside the dislocation channels as well as the only basal slip activation observed for specific mechanical tests. From this analysis, the first polycrystalline model is developed for irradiated zirconium alloys, taking into account the irradiation induced hardening, the intra-granular softening as well as the intra-granular kinematic strain hardening due to the plastic strain localization inside the channels. This physically based model reproduces the mechanical behavior in agreement with the slip systems observed. In addition, this model reproduces the Bauschinger effect observed during low cycle fatigue as well as the cyclic strain softening.     

Bulk-compositional changes of Ni2Al3 and NiAl3 during ion etching

Bulk-compositional changes of Ni₂Al₃ and NiAl₃ in a Ni-50 wt% Al alloy during ion etching have been investigated by transmission electron microscopy and energy dispersive X-ray spectroscopic analyses. After etching with 7, 5 and 3 keV Ar⁺ ions for 15, 24 and 100 h nickel contents in both Ni₂Al₃ and NiAl₃ exceeded greatly those in the initial compounds and increased with the decrement of the sputtering energy. After 100 h etching with 3 keV Ar⁺ ions the compositions of these two compounds reached a similar value, about Ni_80-83Al_12-15Fe_3-4Cr_1-2 (at%). A synergistic action of preferential sputtering, radiation-induced segregation and radiation-enhanced diffusion enables the altered-layers at the top and bottom of the film extend through the whole film. The bulk-compositional changes are proposed to occur in the unsteady-state sputtering regime of ion etching and caused by an insufficient supply of matter in a thin film.     

Mechanical properties of Eurofer 97 in Pb-16Li and irradiation effect

To be used in a fusion reactor, structural materials, and in particular steels, has to be selected and optimised in their composition to achieve a reduction in the long-term radioactive waste. A reduction in the long-term radioactive inventory could be reached substituting elements like molybdenum, niobium and nickel with other ones like tantalum and tungsten which have the same functions as alloying elements and, if irradiated, do not produce long lived radioisotopes. The martensitic steel belonging to the family of 8-9% Cr Eurofer 97 is considered the reference structural steel for fusion application. However, only few information are available about its mechanical properties in the liquid eutectic alloy Pb-16%Li. Particularly, the problem of liquid metal embrittlement (LME) has not been studied in detail and the effect of neutron irradiation on LME has not been investigated at all so far. This work presents the results obtained irradiating tensile specimens of Eurofer 97 up to 5.9 dpa in lead lithium. Tensile tests of samples have been performed out of pile in the same alloy at the same temperature at which irradiation was carried out.     

Irradiation-induced precipitation modelling of ferritic steels

In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Modeling and simulation of irradiation hardening in structural ferritic steels for advanced nuclear reactors

Hardening and embrittlement are controlled by interactions between dislocations and irradiation induced defect clusters. In this work we employ the visco plastic self consistent (VPSC) polycrystalline code in order to model the yield stress dependence in ferritic steels on the irradiation dose. We implement the dispersed barrier hardening model in the VPSC code by introducing a hardening law, function of the strain, to describe the threshold resolved shear stress required to activate dislocations. The size and number density of the defect clusters varies with the irradiation dose in the model. We find that VPSC calculations show excellent agreement with the experimental data set. Such modeling efforts can both reproduce experimental data and also guide future experiments of irradiation hardening.     

Steady-state size distribution of voids in metals under cascade irradiation

The theory of radiation damage in metallic materials predicts that under cascade-irradiation conditions the voids should approach a steady state, which is characterised by a maximum mean void size. It is shown in this paper that the steady-state concentrations of voids of different size are described by the Gaussian distribution with the maximum size mentioned above to be the most probable value. The evolution of voids towards the steady state is analysed.     

Cluster Dynamics modelling of irradiation growth of zirconium single crystals

This paper aims at modelling irradiation growth of zirconium single crystals as a function of neutron fluence. The Cluster Dynamics approach is used, which makes it possible to describe the variation of irradiation microstructure (dislocation loops) with neutron fluence. From the irradiation microstructure, the strain can be calculated along the axes of the lattice structure. The model is applied to the growth of annealed zirconium single crystals at 553 K measured by Carpenter and Rogerson in 1981 and 1987. The model is found to fit the experimentally measured growth of Zr single crystals very nicely, even at large neutron fluence where the ‘breakaway growth’ occurs. This was made possible by considering in the model the growth of vacancy loops in the basal planes. This growth of vacancy loops in the basal planes could be modelled by taking into account that diffusion of self-interstitial atoms (SIA) is anisotropic and that there exist in the basal planes some nucleation sites for vacancy loops (iron clusters), the density of which is considered constant over time.     

Investigation by slow positron beam of defects in CLAM steel induced by helium and hydrogen implantation

Hydrogen and helium ion beams delivering different doses are used in the ion implantation, at room temperature, of China Low Activation Martensitic (CLAM) steel and the induced defects studied by Doppler broadening of gamma-rays generated in positron annihilation. Defect profiles are analysed in terms of conventional S and W parameters, measures of relative contributions of low and high-momentum electrons in the annihilation peak, as functions of incident positron energies E up to 30 keV. The behaviours of the S-E, W-E and S-W plots under different implantation doses indicate clearly that the induced defect size has obvious variation with depth, taking values that interpolate between surface and bulk values, and depend mainly on helium ion fluences. The S-W plot indicates that two types of defects have formed after ion implantation.     

Atomistic simulations of the elastic properties of helium bubble embedded aluminum

The helium bubble has significant consequence to the mechanical properties of irradiated materials. The influence of embedded helium bubble to the elastic properties of aluminum has been investigated by molecular dynamics (MD) simulations. The interaction between aluminum atoms and the interaction between helium atoms are described by an embedded-atom-method (EAM) many-body potential and a pair potential, respectively. Another pair potential, which is parameterized based on ab initio calculation, is used to describe the interaction between aluminum and helium atoms, and its validation under pressure up to 10 GPa is reasonable demonstrated by the electron density calculation. For the composite system consisting of 62,500 aluminum atoms and one helium bubble with various diameters, its elastic constants are calculated properly by stress-strain relation rather than by energy-strain relation. The results show that elastic constants c₁₁, c₁₂ and c₄₄ decrease with increasing of the volume of the helium bubble, and remain almost invariable with the internal pressure of the helium bubble. The main reason is under high-pressure the helium is softer than aluminum, and the soft effect overwhelms the hard effect of internal pressure of helium bubble.     

Microprobe analysis of neutron irradiated and autoclaved zirconium niobium claddings using synchrotron-based hard X-ray imaging and spectroscopy

Two zirconium alloys (Zr-2.5%Nb) - one oxidized in a pressurized water reactor, the other oxidized in autoclave and used as reference - are analyzed by combining synchrotron-based scanning transmission and fluorescence X-ray microscopy and micro-X-ray absorption spectroscopy (micro-XAS). Two-dimensional zirconium distribution maps recorded on the neutron irradiated and the non-irradiated autoclaved Zr-2.5%Nb alloys clearly allow the localization of the oxide and the metal parts of the interface with a micrometer spatial resolution. Micro-XAS investigations make possible the determination of the speciation of zirconium and niobium both in the oxide and the metal parts of the interface for the irradiated and non-irradiated samples. The coordination environment and/or the valency of zirconium and niobium in the metal and the oxide parts of the interface have been determined for both materials, and interpreted on the basis of comparison with metal and oxide reference compounds.     

Radiation flux and damage at accelerator-driven spallation neutron sources

Radiation damage (displacement, helium, and hydrogen production) at proton-driven spallation neutron sources is analyzed and compared for SNS SB (316SS at the nose of the Hg-container vessel), SNS PEW (Al6061 at a hypothetical proton entrance window), and SINQ EW (Al-3 wt% Mg entrance window at Target 5). Spallation neutrons at the three components exhibit differential fluxes, ?′, that increase monotonically with decreasing energy E. For SINQ EW, ?’ is roughly proportional to 1/E, which is attributed to the moderating effect of the D₂O coolant and moderator tank. For 316SS at SNS SB, the calculated total displacement production rate due to protons and neutrons is 34 dpa/yr at full power, with about 37% due to protons. For the Al at SNS PEW and SINQ EW, however, the total rate is 4-5 dpa/yr, with about 90% due to protons. He and H production in all three components is dominated by the incident protons. For He, comparison of experimental and calculated production cross sections for protons on 316SS and Al indicates the need to employ the non-default Jülich ILVDEN option in running LAHET. The resulting total production rates for SNS SB, SNS PEW, and SINQ EW are about 3000, 2400, and 1900 appmHe/yr, respectively. These rates are 1.5-2 times the rates previously calculated using the default GCCI ILVDEN option. The high mobility of H atoms promotes H escape from thin targets of 316SS and Al. For 0.1 cm-thick samples, we tallied the H where it comes to rest using IOPT 14, and obtained production rates at SNS SB, SNS PEW, and SINQ EW of 11500, 4300, and 3500 appmH/yr, respectively.     

Amorphisation of ZnAl2O4 spinel under heavy ion irradiation

ZnAl₂O₄ spinels have been irradiated with several ions (Ne, S, Kr and Xe) at the IRRSUD beamline of the GANIL facility, in order to determine irradiation conditions (stopping power, fluence) for amorphisation. We observed by transmission electron microscopy (TEM) that with Xe ions at 92 MeV, individual ion tracks are still crystalline, whereas an amorphisation starts below a fluence of 5 × 10¹² cm⁻² up to a total amorphisation between 1 × 10¹³ and 1 × 10¹⁴ cm⁻². The coexistence of amorphous and crystalline domains in the same pristine grain is clearly visible in the TEM images. All the crystalline domains remain close to the same orientation as the original grain. According to TEM and X-ray Diffraction (XRD) results, the stopping power threshold for amorphisation is between 9 and 12 keV nm⁻¹.     

A kinetic Monte Carlo approach for the analysis of trapping effect on the defect accumulation in neutron-irradiated Fe

The trapping effect of self-interstitial atom (SIA) clusters in neuron-irradiated Fe was analyzed in terms of generic traps. The effect of the cut-off size between sessile and glissile SIA clusters was investigated. The accumulation of SIA clusters decreased drastically as the cut-off size increased, which originated from the elimination of the SIA clusters at a grain boundary through its one-dimensional motion. When the immobile generic traps were introduced to the kinetic Monte Carlo simulation model, the effect of trap parameters was assessed. An increase in the binding energy between the trap and SIA-species resulted in a decrease in the number of mono-SIAs that were dissociated from the trap and a corresponding delay in visible SIA clusters. The size-dependent prefactor for the dissociation rate of trapped SIA clusters was necessary for a realistic accumulation behavior of SIA clusters. The trap density affects the density and size of the accumulated SIA cluster density during irradiation. This parameterization of generic traps provided insight into the mechanism of accumulation of SIA and SIA cluster.