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Summary

One of the critical problems that affect the cracking of light water reactors core components is the Irradiation-Assisted Stress corrosion cracking (IASCC). Austenitic stainless steel was irradiated in reactors so that studies on IASCC could be conducted. Irradiated microstructure and the role of microchemistry in IASCC were investigated in this study. The dislocation loop microstructure and radiation hardening mechanism, radiation-induced segregation at grain boundaries mechanism, and deformation mechanism have been used to explain IASCC.

This study was almost similar to a research conducted by Paul on stress corrosion cracking (SCC) in stainless steel; however, this was conducted in chlorine and fluorine environments. The findings indicated that radiation had little or no effect on stainless steel corrosion in aqueous solutions. Electrochemical theory and stress sorption cracking theory was used to explain SCC.

The influence of neutron irradiation on stainless steel components inside a reactor vessel leads to changes during plant operations. The radiation primarily induces new point defects. The defects migrate, interact with the original microstructure in relation to the radiation temperature, and stable complexes of the defects are formed. The stainless steel undergoes processes like swelling, toughness, losses ductility, and hardens, which are defects of microstructure, as well as undergoes radiation-induced segregation.

The studies conducted earlier used spectroscopic techniques for microchemical analysis for characterization of grain boundary. Oxygen and fluorine contamination led to intergranular SCC. The failure of type 304L and 304 stainless steel core was explained by the oxyflourine-assisted stress corrosion cracking. Chromium, fluorine, and oxygen strongly influenced intergranular stress corrosion cracking. A recent study investigated the effects of large amounts of helium in a reactor on IASCC  susceptibility. The results indicated that with no significant displacement damage, helium does not lead to IASCC and intergranular cracking in hydrogenated high-temperature water and high-temperature air, respectively. The susceptibility for IASCC and irradiated microstructure was determined recently for neutron-irradiated cold–worked 316 stainless steel with various damage levels. IASCC threshold stress was 45% of the irradiated yield strength for neutron-irradiated steel between 46.9 dpa and 125.4 dpa.

The various studies conducted have illustrated that increase in dose affects the degree of intergranular stress corrosion cracking. Hardening, segregation, voids, and dislocation loops are various ways through which the microstructure of alloys keeps changing as they all correlate with increased susceptibility to cracking. In light water reactors (LWR), the irradiated microstructure is usually dominated by the evolution and formation of faulted dislocation loops. This leads to increased yield strength and hardness and a reduction in uniform elongation. 

Radiation-Induced Segregation

The changes that are related to yield strength may cause effects in the cracking of austenitic alloys. The crack growth rate as well increases with an increase in yield strength in stainless steel in 288 ? BMR water. The occurrence of IASCC is contributed greatly by radiation induces segregation. The irradiated microstructure may be affected by stacking fault energy which varies depending on the composition of alloys. The determination of IASCC is challenged by the fact that localized deformation, hardening, loop microstructure, and RIS tend to saturate at similar doses, and they develop at almost the same rates. At relatively low doses, the IASCC susceptibility is dominated by mechanisms like hardening, relaxation, RIS, and radiolysis.  The amount of cracking and IASCC increases with exposure to neutron. This is a study that is based on the effects of microchemistry and irradiated microstructure on IASCC.

Irradiated-assisted stress cracking (IASCC) is the cracking propensity of an alloy in the presence of irradiation in water that has a high temperature. Reduced time for initiation of crack or increase in growth of a crack compared to its unirradiated counterpart indicate the accelerated cracking. The water chemistry and material are both affected by radiation. Physical changes to microstructure like hardening and segregation that are radiation-induced are the effects of Irradiation in austenitic alloy. Irradiation hardening and radiation-induced segregation cause IASCC. The authors expound on the factors that cause IASCC as outlined in the segments below. 

The non-equilibrium process occurring at point defect sinks during alloy irradiation at intermediate temperature is known as radiation-induced segregation (RIS). The unequal participation of solutes in vacancy and interstitial fluxes to sink results from segregation that is induced by solutes. Depleting the element at the defect sink or enrichment of an alloy element occurs when there is unequal participation of solutes. The association of the interstitial flux with the solute also causes depletion and enrichment of elements at the defect sink. Oversized elements deplete, whereas enrichment occurs to undersized elements. The segregation in Fe-Cr-Ni alloys can be explained by the inverse Kendal mechanism or vacancy exchange. A net solution flux away from or toward the boundary results from an alloying element with vacancy flux. The exchange of a solute with vacancy flux is one of the significant contributors of RIS in Fe-Cr-Ni alloys. The interaction of undersized atoms with interstitial flux is another contributor of RIS in austenitic alloys like Fe-Cr-Ni. The chromium is depleted at the grain boundaries while nickel is enriched, whereas iron can deplete or enrich as it's depended on the composition of bulk alloy. The experimental results below show the RIS of silicon, nickel, chromium, and phosphorous at the grain boundary. The stainless steel was irradiated in a water reactor to several dpa at 300?.

Irradiated Microstructure

showing RIS of chromium

Figure 1: showing RIS of chromium, silicon, phosphorous, and nickel at the grain boundary of irradiated stainless steel in a water reactor.

At LWR service temperatures, the microstructure of austenitic alloys like stainless steel changes rapidly. At very low doses, the point defect clusters are formed. Network dislocation densities and dislocation loops evolve over several dpa with dose. This was confirmed by an experiment conducted on 7 alloys that were irradiated to one and five dpa at 288 ?. The voids and dislocation loops varied over the irradiation doses and alloys. A possibility for formation of voids, He filled bubbles and precipitates in core components that have been exposed to higher temperatures and dose. During radiation, the collapse of the damage cascade may lead to the formation of defect clusters that may be of interstitial or vacancy type. The high mobility of interstitials leads to the growth and nucleation of large dislocated loops. The number density and loop population will grow until the population saturates when interstitials and absorption of vacancies equalize. At 1 dpa, the saturation of loop number density occurs, whereas the loop size evolves to 3-5dpa.

The loop size and number density depend on alloying elements and conditions of irradiation. The number densities are in the order of 1×10 m- and the loop size hardly exceeds 20nm. This was contradicted by an experiment conducted to determine the effect of alloy chemistry on loop size and density. The findings indicated no observed correlations between alloy chemistry and loop size and density. However, the varying experimental conditions may have suppressed the correlations. The reduction of loops as a defect sink helps bubbles and voids grow. The dislocation structure becomes the dominant microstructure as the loop size and number density increase with temperature. Larger Frank loops unfault is made from the dislocation structure that evolves into a network structure. Frank loops are primarily a contribution to radiation hardening compared to network dislocations and black spots at relevant temperatures of LWR.

The growth of second phases can be retarded or accelerated as well as new phases can be produced, and existing phases can be modified by irradiation. The radiation-induced segments of minor and major elements can induce phase formations exceeding the limit of solubility. However, the effect of stability and phase formation by radiation occurs above 400 ?. Irradiation at higher temperatures can lead to induction of lava phase, γ-NiSi, and G phase formation. This was confirmed by an experiment that predicted second phase molecules at 400 ?. The results obtained indicated that precipitation of G phases and γ' was enhanced by dislocation density after irradiation.

showing a simulated evolution

Figure 2: showing a simulated evolution of precipitate in 316SS as a function of aging time at 400?.

Austenitic alloys like stainless steel are essential in structural materials. Irradiation hardening is a great concern for reactor terminals used for long-term services and is placed in environments exposed to radiation. Radiation hardening results from dislocation loop microstructure. The dislocation network and loop react elastically upon the application of stress, thus increasing the yield strength of the austenitic alloy. Indication hardness measurement or tensile tests are used to determine the increase in yield strength. In cases of proton radiation, implantation of almost 3×10 produced gas bubbles and vacancy complexes that gave the dislocation motion a stronger barrier causing hardening. The hardness obtained from Xe and H are studied, and protons produce a steep peak displacement damage that is non-uniform. The saturation fluence of a proton is higher. 

showing different hardness saturation obtained for Xe and proton radiation. 

Figure 3: showing different hardness saturation obtained for Xe and proton radiation. 

Deformation mode changes dramatically with increasing dose in addition to hardening. Narrow channels which have been cleared of defects by preceding dislocations are localized plasticity. The sharp reduction in uniform elongation and localized necking are caused by intense shear bands that result from dislocation channeling.   

The IASCC of irradiated alloys is dependent on similar parameters as the IGSCC of unirradiated alloys. Changes to the alloy and changes in water chemistry are factors through which radiation affects IG cracking.

The origin of radiation effects on water chemistry is represented by the formation of reductants and oxidants. The thermodynamics and electrochemical reactions are controlled by corrosion potential, which is influenced by reductants and oxidants that are formed as a result of radiation. A sigmoidal decrease in corrosion potential is observed when mass transport is limited by the reduction in oxidant levels. Materials pre-irradiated by slow strain rate were studied. A decrease in average crack growth rate was observed with decreasing potential of corrosion. Corrosion potential is the dominating factor in unirradiated alloys, and IGSCC of irradiated that is controlled by ionic species and oxygen content in water. This is confirmed by Anna by highlighting that water dissociates into different radical products of reaction, ionic and molecular products under the influence of radiation. The radical reaction products react to form O2, H2, and hydrogen peroxide (H2O2). The products of radiolysis are scavenged by the addition of hydrogen to reactor water hence reducing the effect of radiolysis. In BMR, the products of dissociation of water lead to an increased potential of corrosion.

The mechanisms of IASCC are categorized into effects of material and water chemistry. The principal effects of material and water chemistry indicate a steep increase of kinetics of environmental cracking with an increase in corrosion potential. There is a sharp increase in the growth rate of crack for a solution with conductivities that are above 0.1 µS/cm in unirradiated or irradiated cases. Solution conductivity and corrosion potential in constant water chemistry conditions control the cracking rate3. IASCC is strongly influenced by the cumulative radiation effect, which is illustrated by the increase in cracking susceptibility with fluence in rare tests and slow strain.

IGSCC susceptibility in alloys like stainless steel and grain boundary chromium level show a correlation where thermal sensitization causes grain boundary depletion. In austenitic alloys, chromium depletion is an agent in IGSCC in oxidizing conditions. The boundary becomes a location for oxide destabilization, and passivation by the formation of a protective chromium oxide over the grain boundary is not possible when the concentration of chromium grain boundary drops below the threshold level. The chromium depletion profiles that result from precipitation reactions or RIS show a difference by having a smaller width of RIS profiles by two orders of magnitude. Chromium depletion at the grain boundary is insufficient but an important condition for the occurrence of IASCC. However, the depletion of chromium in oxygenated water was seen to be sufficient to cause IASCC. This is because post-irradiation annealing showed a change in IG cracking susceptibility in BMR water while RIS was virtually unchanged. RIS is found to be dependent on GB structure, the dose of irradiation, and the composition of alloy. The structure of GB has a strong ability to suppress segregation to GBs under irradiation.

The dislocation of loop microstructure on alloy behavior is hardening which follows a dependence on the square root of the product of loop size and loop number density according to the hardening model of the dispersed barrier. Hardening is an essential factor in the susceptibility of IASCC. Increased cold work leads to an increase of crack growth in austenitic stainless steel that is not sensitized. Increased yield strength from cold work can predict the susceptibility of cracking. This was confirmed by a study that found that cracking susceptibility is promoted by cold-working and subsequent deformation during SCC tests.

The parallel between IASCC irradiation and cold work effects indicates that hardening is an essential factor in IASCC susceptibility. Cracking and hardening are reduced by increased annealing, but the behavior of hardening cannot satisfactorily explain the complete and rapid reversal of susceptibility in IASCC with annealing condition 3. The susceptibility of cracking changes drastically before the dropping of hardness from its irradiated value begins. The various set of experiments that were conducted by Hash revealed that hardening was not the only factor in IASCC. This is because IASCC susceptibility was not constant when hardness was the only sole factor. Cracking was also higher in lower doses than in companion samples that had similar doses with no cold work. This indicates that IASCC susceptibility can also be enhanced by cold work.

The IASCC proton irradiated samples over a wide range of doses have shown that nickel alloys have high IASCC resistance. The hardening and dislocation microstructure is similar in 18 chromium- 8 nickel alloys, and through a dose of 5 dpa, no voids were observed. The higher content of nickel in stainless steel increases significantly the stacking fault energy (SFE). A change in slip nature from a planar slip to a wavy slip results from increasing SFE. The SCC resistance in stainless steel has been linked to SFE. In stainless steel, decreased SCC susceptibility and increased area reduction correlated well with increasing SFE. The rupture of an oxide could be caused by stress at grain boundary at lower fluences before the occurrence of matrix material cracking leading to subsequent corrosion, exposure of the material to the solution, and IASCC.

IASCC can also occur by deformation of the boundary leading to oxide film rapture causing IASCC. Localized deformation is needed to break oxide film that covers grain boundary. The array of dislocations irradiation causes the formation of dislocation channels. The formed dislocation channels terminate or transmit at the grain boundary. A greater impact on dislocation channels results when larger dislocations are involved since the number of glide dislocations in the channel is proportional to the height of the channel.

The experiments conducted indicated that cold work was less effective than radiation hardening in the induction of IGSCC. Slip bands and dislocation channels can influence cracking. The grain boundary stacking fault energy (GB SFE) slightly increases with doses for all alloys as there was no clear difference in dose dependence for any alloy. The GB SFE is increased by nickel enrichment from RIS, whereas molybdenum and chromium depletion lowers the value. The experimental results obtained indicated that lower SFE is found in susceptible 304 alloys than 316 alloys that are resistant in the unirradiated condition. This suggests that SFE at the grain boundary could have an influence on the determination of IASCC susceptibility.

The authors were able to identify that the inverse Kendal mechanism or vacancy exchange could be used to explain radiation-induced segregation in Fe-Cr-Ni alloys. This was confirmed by research that also identified that solute exchange with vacancy flux was a significant contributor of RIS in Fe-Cr-Ni alloys, as well as the interaction of interstitial flux with undersized atoms. The study also highlighted that point defect clusters are formed at very low doses because, at LWR service temperatures, the microstructure of austenitic alloys changes rapidly. This was verified by a study conducted on 7 alloys that were irradiated to 1 and 5 dpa at 288 ?. The voids and dislocation loops varied over irradiated doses and alloys. The study also highlighted that increment in yield test can be used to indicate hardness measurements. This as well is authenticated by radiation hardening results that were obtained from the dislocation of loop microstructure. The dislocation loop and network reacted elastically upon the application of stress; hence the yield strength of the austenitic alloy increased.

Ionic species and oxygen content in water control IGSCC of irradiated alloys observation was confirmed by Anna, who highlighted that water forms radical products upon irradiation that react to form O2, H2, and hydrogen peroxide (H2O2), increasing the potential of corrosion. Chromium depletion was identified as an agent in IGSCCin oxidizing conditions. Research by Onchi et al confirmed that chromium depletion in oxygenated water was sufficient enough to cause IASCC. The various studies conducted have shown that factors that lead to susceptibility of IASCC are dependent on each other, as the authors had indicated.

This study is limited because localized deformation, hardening, loop microstructure, and RIS tend to saturate at similar doses and develop at almost the same rates. Therefore, the occurrence of IASCC is dependent on factors that correlate with each other. The primary factor causing susceptibility of IASCC was not identified. The SFE, deformation of the boundary leading to the rapture of oxide film, cold work, and hardening all contribute to the susceptibility of IASCC. 

Conclusion

The change in hardness of alloys like stainless steel, microchemistry, and microstructure are resultant from irradiation of stainless steel. The alloy composition and microstructure that has been irradiated can affect the mode of deformation compared to a sample that has not been irradiated. Deformation of the material at the grain boundary through the oxide film rapture is a potential mechanism for irradiation-assisted stress corrosion cracking. Boundary dislocations result in sliding of the grain boundary or localized slip that causes oxide film rapture hence initiating and propagating IG cracks. The radiation-induced segregation alters SFE leading to the enhancement or diminishing of the planarity slip, thus causing changes at the grain boundary. The RIS is found to be dependent on factors like GB structure, the composition of the alloy, and the dose of irradiation. The structure of GB has the highest capability for suppression of segregation to GBs under irradiation. The dependency of loop size and number density on conditions of irradiations and alloying elements was contradicted by a study conducted on chromium and nickel. The results obtained indicated no observed correlation between alloy chemistry and loop size or density. The primary cause for ISACC has not been identified as it is depended on both the way deformation of a material occurs and the state of the material that is irradiated. The rapture of oxide films near the GB is a potential ISACC mechanism.

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