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Gradient-tempering process increases fracture toughness and resistance to stress-corrosion cracking of ball-bearing races made of hard, strong steels and subject to high installation stresses and operation in corrosive media. Also used in other applications in which local toughening of high-strength/low-toughness materials required.
Since the late 1990’s, the Hagan container was used as the primary container for packaging of plutonium-bearing materials. The Hagan design consisted of a threaded closure, a Viton® ORing, a carbon-carbon filter, and a 304L stainless steel (SS) body. Over the years, Hagans have shown vulnerability in their design [1]. In 2008, The Department of Energy (DOE) issued DOE M 441.1-1, Nuclear Material Packaging Manual, which detailed an approach to obtain highconfidence in containers by including specific design requirements, material contents and an approach to determine life span from said contents, and surveillance techniques [2]. In response to both the vulnerability issues with the Hagan and DOE M 441.1-1, the SAVY 4000 container with its twist style closure, Viton® O-Ring, Fiberfrax-Gortex filter, and annealed 316L SS body, was designed as the replacement for Hagan containers, but only for a short term lifespan of 5 years [1]. However, both the Hagan and SAVY 4000 are being pushed to maintain a lifespan of 40 years. Therefore, proper confidence must be placed on each component of each container to last a minimum of 40 years. So far, the biggest concern found during surveillance of these containers is corrosion and the potential for failure by corrosion. One concern is that the containers fail due to stress corrosion cracking (SCC), especially around the weld between the collar and the body as welds leave residual stresses. One advantage the SAVY 4000 has is that the body is annealed, but its weld is still susceptible as it was welded after annealing [3, 4]. Moreover, 316L SS is known to have a higher pitting resistance (pits are a precursor to SCC and can also lead to extensive failure of the material), than 304L SS [4]. During recent surveillance activities, two SAVY 4000’s containing Solution Assay Instrument (SAI) solutions were opened. The SAI SAVY 4000’s contained plutonium (Pu) in 3M HCl solution in plastic volumetric flasks placed inside of polyethylene bags. Historically, a SAVY 4000 containing an SAI solution is packaged for 3 weeks, however, these particular containers were not reopened until 14 months later. When opened, brown-red corrosion product was found all over the inside of the container as shown in Figure 1.
Corrosion has become a decisive factor hindering supercritical water technologies' long-cycle, safe implementation. Electrochemical experiments performed in supercritical aqueous systems can capture essential, real-time information about the micro/nanoscale processes of corrosion and their kinetics. As a starting point, this paper first summarizes the critical chemical properties of supercritical aqueous systems and analyzes their role in determining the duality of corrosion mechanisms (whether electrochemical or chemical micro-processes predominate). Next, it reviews the available electrochemical techniques such as pH sensors, electrochemical noise measurements, and related applications for quantitatively elaborating the corrosion issues in supercritical aqueous systems. With thick and heavy colors, a series of mechanistic models, such as the mixed potential model and the coupled environment models were introduced from the fundamental theories to the related applications, which have been applied widely in ambient and subcritical aqueous systems and exhibit great values and potential to be employed in high-density supercritical water environments. The passivity, depassivation, and passivity breakdown of metals and alloys in terms of the Point Defect Model (PDM) were also successfully implemented and discussed. Further, the newly built SCW_PDM describes the film growth kinetics at the atomic level of metals and alloys in supercritical water and is being developed to quantitatively analyze and resolve a series of corrosion issues such as the passivity breakdown and the stress corrosion cracking that occur in supercritical aqueous systems. This review is helpful by promoting the development of the electrochemical corrosion theory in supercritical aqueous systems and by identifying corrosion prevention and control methods for the safe implementation of relevant supercritical water technologies.
Austenitic stainless steel welds play a crucial role in maintaining the structural integrity of light water reactors. Within the reactor core, stainless steel welds endure a harsh service environment including elevated temperature, corrosive coolant, and intense neutron irradiation. Long-term exposure to such a severe service environment could lead to performance degradation. To ensure the safe and economic operation of light water reactors, this study performed crack growth rate and fracture toughness tests on irradiated and thermally aged weld samples harvested from decommissioned reactor components. In a low-corrosion-potential environment, stainless steel welds showed resistance to stress corrosion cracking, and the measured crack propagation rates were comparable to those of wrought stainless steels under similar test conditions. However, a marked decline in fracture toughness was observed in weld samples, particularly in irradiated samples. Fracture instability was observed at ~0.8 dpa, significantly lower than those observed in wrought stainless steels. Further research at higher doses is needed to evaluate the dose dependence of irradiation-induced embrittlement, and to understand the interplay between thermal aging and neutron irradiation in stainless steel welds.
Cold working and double aging yield high strength without sacrifice of resistance to corrosion. Report presents data on mechanical properties and stress-corrosion resistance of triple-melted, solution-treated, work-strengthened, direct-double-aged Inconel(R) 718 alloy. Triple melting consists of vacuum induction melting, electro-slag remelting, and vacuum arm remelting. Data indicate advance in processing of large-diameter bars. New process increases yield strength without reducing the elongation, reduction of area, and grain size.
The continued operation of aging Boiling Water Reactors (BWRs) worldwide requires gradually increasing cost of inspection, maintenance, and repair. Intergranular Stress Corrosion Cracking (IGSCC) in sensitized austenitic stainless steel piping first became a major issue for BWRs in the 1980s, resulting in recognition of the susceptibility of reactor internals to IGSCC. Shroud cracking identified in 1993–1994 confirmed that IGSCC of internals is a significant issue for BWRs. IGSCC is a time-dependent, material degradation process, which is caused and accelerated by the presence of residual stresses, material sensitization, irradiation, cold work, elevated temperature, and corrosive environments. This paper emphasizes the importance of accounting for corrosive environments, or more exactly, electrochemical phenomena in modelling IGSCC and predicting the service life of BWR in-vessel components, and stresses the necessity of performing such modeling not just for a single state point under full power conditions, but for the whole operating history of the reactor, including startups and shutdowns. Overall, this paper demonstrates that ignoring electrochemical considerations may result in underestimating component lifetime and lead to unnecessary expenses for inspection and repair.
This report is a critical review of both environment-enhanced fatigue crack propagation data and the predictive capabilities of crack growth rate models. This information provides the necessary foundation for incorporating environmental effects in NASA FLAGRO and will better enable predictions of aerospace component fatigue lives. The review presents extensive literature data on 'stress corrosion cracking and corrosion fatigue.' The linear elastic fracture mechanics approach, based on stress intensity range (Delta(K)) similitude with microscopic crack propagation threshold and growth rates, provides a basis for these data. Results are presented showing enhanced growth rates for gases (viz., H2 and H2O) and electrolytes (e.g. NaCl and H2O) in aerospace alloys including: C-Mn and heat treated alloy steels, aluminum alloys, nickel-based superalloys, and titanium alloys. Environment causes purely time-dependent accelerated fatigue crack growth above the monotonic load cracking threshold (KIEAC) and promotes cycle-time dependent cracking below (KIEAC). These phenomenon are discussed in terms of hydrogen embrittlement, dissolution, and film rupture crack tip damage mechanisms.
The performance of structural materials is critical for the safe and economic operation of light water reactors. Exposed to neutron irradiation during service, the reactor core internal materials can undergo significant microstructural and microchemical changes, leading to irradiation hardening and embrittlement. To ensure the structural integrity and functionality of nuclear reactor components during long-term operation, material degradation and damage mechanisms must be understood and managed adequately. In this work, irradiated materials harvested from the decommissioned Zorita reactor were studied for their cracking susceptibility and fracture resistance as a function of irradiation dose up to 47 displacement per atom (dpa). The material is a Type 304 stainless steel sectioned from the baffle plates of this pressurized water reactor with 38 years of service. Crack growth rate and fracture toughness J-resistance (J-R) curve tests were performed in low-corrosion-potential environments at ~315°C. All samples behaved similarly under cyclic loading, and no deteriorated corrosion-fatigue behavior was observed in the test environments. Under constant stress intensity factors, most samples did not show elevated crack growth rates, suggesting an adequate stress corrosion cracking resistance in the test environments. However, an unstable cracking behavior was observed in a 47-dpa sample, resulting in significantly higher crack growth rates than expected at high stress intensity factors. Crack instability was also observed in a 0.06-dpa sample but did not lead to a sustained high crack growth rate. The impact of neutron irradiation was more evident in the fracture toughness J-R curve tests. As the dose increased, the J-R curve declined considerably and became very shallow at high doses. A fully intergranular fracture morphology was also observed among the high-dose samples ruptured in an air atmosphere at room temperature. This brittle fracture mode in the absence of high temperature water environment confirmed a high degree of embrittlement of this material resulting from its service exposure to neutron irradiation.
Ni-based structural alloys in molten salt environments often experience simultaneous mechanical loading and corrosive attack, yet the mechanisms governing stress-corrosion interactions remain unclear. Prior studies largely emphasize tensile stress, while the role of compressive stress has received limited attention. Here, reactive molecular dynamics simulations are used to investigate the coupled effects of applied strain and corrosion in Ni 0.75 Cr 0.25 exposed to molten FLiNaK at 800 °C. A Σ5(210) grain boundary model is subjected to tensile (+4%) to compressive (−4%) uniaxial strains, and corrosion behavior is evaluated through fluorine adsorption, charge redistribution, and grain boundary evolution. Tensile strain accelerates intergranular corrosion susceptibility by reducing local atomic packing through elastic dilation and increasing excess free volume at the grain boundary, which enhances atomic mobility and salt infiltration. In contrast, compressive strain can suppress corrosion by promoting the formation of a ridge-like surface layer along the grain boundary, limiting salt access to the underlying alloy. These results provide atomistic insight into how stress states influence grain boundary corrosion in molten salts.
A dissimilar metal weldment consisting of SA508-309L-308L-304L is widely used in light-water nuclear reactors. These weldments demonstrate dissimilar susceptibility to stress corrosion cracking that are related to the microstructure, properties, and residual stress. In this work, microstructures, hardness, and the residual stress distribution of the dissimilar metal weldments were investigated, with the correlation of increased hardness in the heat-affected zone (HAZ) to the microstructure. 304L HAZ demonstrated similar grain morphology as the base material, and the increase in hardness was primarily attributed to the increased dislocation density. SA508 HAZ demonstrated a change of grain morphology resulting from the different peak temperatures and cooling rates. The increased hardness in the SA508 HAZ was attributed to the refined grain morphology, higher dislocation density, and higher number density of precipitates. A ~ 20–30-μm-wide martensitic zone formed at the fusion boundary of SA508-309L, where Cr-rich carbide precipitates were observed, with the average size and the number density of 44.1 ± 16.9 nm and 1.5 × 10 21 m ₋3 , respectively. In conclusion, residual stress results demonstrated the largest tensile stress at 309L butter, indicating its high cracking susceptibility.
Weld residual stress is one of the primary driving forces for primary water stress corrosion cracking in dissimilar metal welds (DMWs). To mitigate tensile residual stress in DMWs, it is critical to understand residual stress distribution by modeling techniques. Recent studies have shown that weld residual stress prediction using today s DMW residual stress models strongly depends on the strain-hardening constitutive model chosen. The commonly used strain-hardening models (isotropic, kinematic, and mixed) are all time-independent and inadequate to account for the time-dependent (viscous) plastic deformation at the elevated temperatures experienced during welding. For materials with profound strain-hardening, such as stainless steels and nickel-based alloys that are widely used in nuclear reactor and piping systems, the equivalent plastic strain the determinate factor of the flow stress can be highly dependent on the recovery and recrystallization processes. These processes are in turn a strong function of temperature, time, and deformation rate. Recently, the authors proposed a new temperature- and time-dependent strain-hardening constitutive model: the dynamic strain-hardening constitutive model. The application of such a model has resulted in improved weld residual stress prediction compared to the residual stress measurement results from the contour and deep-hole drilling methods. In this study, the dynamic strain-hardening behavior of Type 304 stainless steel and Alloy 82 used in pressure vessel nozzle DMWs is experimentally determined. The kinetics of the recovery and recrystallization of flow stress are derived from experiments, resulting in a semi-empirical equation as a function of pre-strain, time, and temperature that can be used for weld residual stress modeling. The method used in this work also provides an approach to study the kinetics of recovery and recrystallization of other materials with significant strain-hardening.
The proposed project aims to develop a friction sir based repair technique to heal cracks of stainless steel dry storage canisters (DSCs), created by the effect of stress corrosion cracking (SCC). In real service conditions, the welded parts of the canisters become prone to SCC under exposure to aggressive chemical environment. The focus of the project is on developing a mitigation strategy that can be implemented for countering SCC in DSCs following a two-pronged approach: (i) apply friction stir processing (FSP) as a crack repair technique while creating compressive residual stresses at the canister material surface, thus improving resistance against SCC; (ii) add molybdenum and/or nitrogen by friction stir based alloying process to create compositions that improve pitting resistance and thus enhance resistance against SCC. The specific project objectives are listed below: • Investigate FSW of 304L SS a potential repair welding technique of dry storage canisters; • Optimize the FSW parameters to introduce surface compressive stress profiles in the weldments; • Develop a FSP method to surface alloy the 304L with nitrogen and Mo by adding CrN and Mo powders; • Evaluate the surface compressive residual stress profile during FSW and FSP processes; • Evaluate the localized corrosion resistance of the FSW and FSP samples as a function of temperature, chloride concentration, and surface alloy composition; • Evaluate the SCC behavior of the FSW and FSP samples as a function of stress, chloride concentration, and temperature at different surface chemistries and induced residual stress profiles.
A preliminary study on the microstructural characteristics and stress corrosion cracking susceptibility of a friction stir welded (FSW) 304L stainless steel plate was carried out. The weld examined was characterized by several typical microstructural features of friction stir welds including a gradient of dynamically recrystallized microstructure with distinct material flow patterns reflective of the complex distribution of thermomechanical histories. Evidence of process-induced microstructural sensitization was lacking Immersion testing of the friction stir welded plate in boiling magnesium chloride solution indicated the FSW region was more susceptible to SCC than the base 304L material, especially along the weld toes. The microstructural origins of this SCC susceptibility are not clear, but it is likely driven by residual stress imparted by the welding process. Future work will focus on direct examination of the SCC damaged microstructure and residual stress of the weld zone to further clarify the operative characteristics controlling SCC susceptibility.
The fundamental principles allowing design of stainless bearing steels with enhanced toughness and stress corrosion resistance has involved both investigation of basic phenomena in model alloys and evaluation of a prototype bearing steel based on a conceptual design exercise. Progress in model studies has included a scanning Auger microprobe (SAM) study of the kinetics of interfacial segregation of embrittling impurities which compete with the kinetics of alloy carbide precipitation in secondary hardening steels. These results can define minimum allowable carbide precipitation rates and/or maximum allowable free impurity contents in these ultrahigh strength steels. Characterization of the prototype bearing steel designed to combine precipitated austenite transformation toughening with secondary hardening shows good agreement between predicted and observed solution treatment response including the nature of the high temperature carbides. An approximate equilibrium constraint applied in the preliminary design calculations to maintain a high martensitic temperature proved inadequate, and the solution treated alloy remained fully austenitic down to liquid nitrogen temperature rather than transforming above 200 C. The alloy can be martensitically transformed by cryogenic deformation, and material so processed will be studied further to test predicted carbide and austenite precipitation behavior. A mechanistically-based martensitic kinetic model was developed and parameters are being evaluated from available kinetic data to allow precise control of martensitic temperatures of high alloy steels in future designs. Preliminary calculations incorporating the prototype stability results suggest that the transformation-toughened secondary-hardening martensitic-stainless design concept is still viable, but may require lowering Cr content to 9 wt. pct. and adding 0.5 to 1.0 wt. pct. Al. An alternative design approach based on strain-induced martensitic transformation during cryogenic forming, thus removing the high martensitic constraint, may permit alloy compositions offering higher fracture roughness.
This paper provides a review of recent developments in the field of Mg corrosion and puts those into context. This includes considerations of corrosion manifestations, material influences, surface treatment, anodization, coatings, inhibition, biodegradable medical applications, stress corrosion cracking, flammability, corrosion mechanisms for HP Mg, critical evaluation of corrosion mechanisms, and concluding remarks. There has been much research recently, and much research continues in this area. In conclusion, this is expected to produce significantly better, more-corrosion-resistant Mg alloys.
A technique has been developed to design a maximum efficiency reliable pressure vessel of given geometry and service life. The technique for ensuring reliability of the minimum weight vessel relies on the application of linear elastic fracture mechanics and fracture mechanics concepts. The resultant design incorporates potential fatigue and stress corrosion crack extension during service of a worst case initial flaw. Maximum stress for safe life is specified by the design technique, thereby minimizing weight. Ratios of pressure and toughness parameters are employed to avoid arbitrary specification of design stress level which would lead to a suboptimum design.
A dynamic fatigue study was performed on a Li2O-Al2O3-SiO2 glass-ceramic in order to assess its susceptibility to delayed failure. Fracture mechanics techniques were used to analyze the results for the purpose of making lifetime predictions for optical elements made from this material. The material has reasonably good resistance (N = 20) to stress corrosion in ambient conditions. Analysis also indicated the elements should survive applied stresses incurred during grinding and polishing operations.
Under PCMI conditions, the cladding strain is driven by either pellet swelling during normal operation or pellet thermal expansion in transients. The loading in PCMI is displacement controlled because friction between the pellet and the cladding results in a strain compatibility at the pellet/cladding interface. Failure is known to occur under PCMI loading conditions during operation when normal power changes result in release of iodine (and other fission products) causing stress corrosion cracking where the crack-propagation rate is dependent upon the cladding stress but slow relative to strictly mechanical failures. Failures after normal power changes are often referred to as failure by pellet-cladding interaction or PCI to distinguish it from PCMI. Such failures are considered displacement-controlled failures, as noted above, and occur with limited plastic strain. This is different to what is typically thought of as PCMI failure under transient conditions, either Anticipated Operational Occurrences (AOOs) or Reactivity Insertion Accidents (RIAs), which is generally assumed to occur once the cladding reaches a mechanical strain limit. But there are two features in the typical PCMI failure in an RIA that are not consistent with the PCMI assumptions. The first is the ductile shear failure feature in cross section and the second is a large crack opening post failure (e.g., see OI-11 failure in NSRR program). The ductile feature is indicative of a plastic instability that requires very large local strains, and the large crack opening must be the result of a release of energy, both of which are not expected if stress in the cladding was only driven by thermal expansion in the pellet. It is concluded that there must be a mechanism (likely related to rod internal pressure) that eliminates strain compatibility between the pellet and cladding – allowing strain localization to occur in what are known as PCMI transient failures. In some sense, these PCMI failures may actually be early burst failures that occur when thermal expansion causes uniform strains greater than a defined limit and fission gas pressure eliminate the strain compatibility