Irradiation Assisted Stress Corrosion Cracking Research Articles

Dislocation channel broadening–A new mechanism to improve irradiation-assisted stress corrosion cracking resistance of additively manufactured 316 L stainless steel

Additively manufactured (AM) 316 L stainless steel (SS) after hot isostatic pressing (HIP) was found to exhibit superior resistance to irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water, as compared to wrought 316 L SS. The well-accepted IASCC factors of radiation-induced segregation (RIS) and radiation hardening are not accurate descriptions of IASCC susceptibility in this case. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was confirmed to suppress crack initiation. A unique distribution of irradiation-induced defects in HIP AM 316 L SS eased dislocation cross-slip compared to those in the wrought counterpart, thus increasing the channel width near the grain boundaries. For the first time, this study highlights the importance of dislocation channel broadening as a potential mechanism to further improve the IASCC resistance of 316 L SS and provides direct experimental evidence based on commercial-grade materials.

Effect of low dose irradiation of heavy ion on electrochemical corrosion and IASCC behavior of austenitic steel

The impacts of low dose irradiation on the behavior of electrochemical corrosion and irradiation assisted stress corrosion cracking for 321 stainless steel were studied using Fe2+ ion irradiation to simulate neutron radiation damage in primary circuit environment of pressurized water reactor. Low dose irradiation can improve the pitting resistance and reduce the cracking tendency of the alloy in B-Li solution to a certain extent, which was related to the δ phase content on the near-surface of the sample: The higher δ phase content on the near-surface of the 2 dpa irradiated sample was observed by grazing incident X-ray diffraction. In addition, the pits was significantly increased near micro-cracks for the unirradiated sample, indicating that the existence of pits induced the initiation of cracks. The research results provided an important reference for the failure mechanism of irradiation assisted stress corrosion cracking of core components.

Irradiation assisted stress corrosion cracking of 347 stainless steel in simulated PWR primary water containing lithium hydroxide or potassium hydroxide

The susceptibility of type 347 stainless steel to irradiation assisted stress corrosion cracking (IASCC) in pressurized water reactor (PWR) primary water with two water chemistries was evaluated. Miniature four-point bend tests were used to assess the effect of potassium hydroxide (KOH) in comparison to lithium hydroxide (LiOH) on the stress to initiate IASCC. Four-point bend specimens were machined from baffle-former bolts removed from the Salem unit 1 and D.C. Cook unit 2 reactors with accumulative doses ranging from 7.8 to 26.4 dpa. It was found that the pseudo-threshold stress to initiate IASCC in this alloy was 50–60% of the irradiated yield strength irrespective of water chemistry. Intergranular cracking was the dominant fracture mode in all specimens and the fracture morphology was characterized by a tortuous fracture path through the oxide and was common to both water chemistries. Grain boundary oxidation combined with localized stress enhancement at deformation band – grain boundary sites were precursors for IASCC initiation in both water chemistries. The pseudo-threshold stress for IASCC initiation is in good agreement with O-ring test results on both 316SS and 304SS. Overall, the stress to initiate cracks in neutron-irradiated 347SS was indistinguishable between simulated primary water containing LiOH or KOH.

Decoupling the roles of grain boundary oxidation and stress in IASCC of neutron-irradiated 304L stainless steel

Irradiation assisted stress corrosion cracking (IASCC) is known to be a combination of applied stress, and a corrosive environment in irradiated materials, but an appropriate understanding of how each of these components affects cracking is not yet clear. In this work, we isolate the role of grain boundary (GB) oxidation in IASCC. For this purpose, solution-annealed 304L stainless steel irradiated in reactor to 5.4 or 69 dpa was studied using a miniaturized four-point bend technique. Two modes of bend tests were conducted; straining in simulated pressurized water reactor primary water (PW) at 320 °C in an incremental manner, and pre-oxidation in PW at 320 °C without application of stress followed by dynamic straining in purified Ar at the same temperature until crack initiation. Exposure of the 5.4 dpa sample for 210 h in high temperature water initiated cracks in Ar at 60% of the yield stress (0.6YS), whereas the companion sample of the same dose exposed for 1010 h cracked in Ar at 0.5YS. The long-term exposure in water led to GB oxidation that ultimately lowered the crack initiation stress. Dynamic straining in water resulted in larger crack lengths and greater crack depths indicating stress accelerated oxidation and cracking. Dislocation pile-ups at dislocation channel-grain boundary sites provide an amplification of the applied stress to a level consistent with the fracture stress of the grain boundary oxide, providing an explanation for IASCC that occurs at applied stresses well below the irradiated yield stress of the alloy. There was no evidence of cracking upon straining of an unoxidized 69 dpa sample in Ar to well above yield, confirming that the irradiated state is not inherently susceptible to intergranular (IG) cracking and that oxidized GBs are responsible for initiation of IG cracking.

Prediction of inter-granular crack initiation susceptibility in post irradiated grade 300 steels

Grade 300 austenitic steel components are subjected to irradiation-assisted stress corrosion cracking hazard, in the context of nuclear pressurized water reactor operation. In this paper, it is assumed that inter-granular cracking susceptibility primarily depends on sub grain plasticity mechanisms, controlling the slip band thickness and spacing achieved for a plastic strain level imposed. A computational model evaluating the related grain-boundary stress evolutions is developed and evaluated, using actual post-irradiation observations. In practice, we calculate a quantitative damage susceptibility factor R depending on the characteristic local slip band arrangements and the grain to grain misorientation. The statistical R evolutions obtained are consistent with the strain localization and related intergranular crack susceptibility trends, under corresponding irradiation and deformation conditions.

Irradiation-assisted cracking of SA508-304L weldments with 308L groove filler and 309L butter in hot water immersion constant rate extension tests

Low alloy SA508 bainitic steel is often welded to 300 series austenitic stainless steel in nuclear power plants. These dissimilar metal weldments use 309L butter applied to the low alloy steel and 308L groove filler material to join the 309L to austenite. High alloy steels (such as the 308L and 309L material used here) have been identified as having greater susceptibility to irradiation-assisted stress corrosion cracking (IASCC). Evidence is presented here of cracking under tensile loads in 309L weldment butter in LWR hydrothermal conditions that correlates irradiation induced changes at δ ferrite−γ austenite phase boundaries and γ−γ grain boundaries; specifically intergranular Cr23C6 precipitates. Cracking is not observed in irradiated 308L groove filler under the same conditions. Both the 309L butter and 308L filler material have a duplex or mixed δ−γ phase microstructure due to solidification during the welding procedure. We have performed constant extension rate autoclave immersion tests in BWR normal water chemistry hot water (288 °C, 10 MPa, 2000 wt. ppb dissolved oxygen, <100 nS/cm conductivity, neutral pH) of proton irradiated (2 MeV, approximately 3 displacements per atom at a depth of 10 μm) tensile specimens. Proton irradiation to the dpa levels used here induced Cr23C6 carbide precipitation at γ−γ grain boundaries and δ−γ phase boundaries in the 309L butter but not the 308L filler. Cracking was observed in the proton-irradiated volume of 309L at these interfaces in the vicinity of the Cr23C6 precipitates. We hypothesis the SA508 is the source of carbon in the 309L butter. Cr depletion at γ−γ grain boundaries was also observed and may also be a root cause of cracking. The 309L butter is under significant tensile residual stress in the weldment that adds to applied tensile loads and this too may be a contributing factor.

Intragranular irradiation-assisted stress corrosion cracking (IASCC) of 316L stainless steel made by laser direct energy deposition additive manufacturing: Delta ferrite-dislocation channel interaction

In this study, irradiation-assisted stress corrosion cracking (IASCC) resistance of as-built AM 316L stainless steel (SS) made by direct energy deposition (DED) was evaluated in a simulated boiling water reactor (BWR) environment. Different from the intergranular cracking typically seen from 316L SS, unique intragranular cracking away from grain boundaries (GB) was observed. The presence of high-density micron-sized delta ferrite near the melt pool boundaries promoted the in-grain strain localization and crack nucleation. Delta ferrite particles can efficiently retard dislocations during deformation, altering the strain localization from the GB to its surrounding area and thus reducing the chance of GB cracking. The phenomenon was further confirmed by the higher nanoindentation hardness measured from the corresponding area, and by characterizing the evolution of microstructure, cracking morphology, and dislocation channels.

Deformation and cracking behaviors of proton-irradiated 308L stainless steel weld metal strained in simulated PWR primary water

The proton-irradiated 308L stainless steel weld metal was strained by using constant extension rate tensile testing in simulated PWR primary water, and its deformation microstructures and irradiation assisted stress corrosion cracking (IASCC) behavior were investigated. The results suggest that the irradiation significantly increases the SCC susceptibility of 308L weld metal and causes various deformation microstructures including lathy faulted planes, dislocation channels and deformation twins in austenite and atomic plane rotation in δ-ferrite. The propagation of intergranular IASCC cracks is closely related to the location of the crack tip. For the crack tip in the specimen matrix interior, localized deformation is likely the key factor responsible for the crack growth. For the crack tip close to the specimen surface, however, localized corrosion along the grain boundary rather than the localized deformation appears to dominate the crack propagation. Unlike the intergranular cracks, the IASCC cracks along the δ-ferrite/austenite phase boundary can initiate either by crack initiation at the phase boundary or by crack propagation from the grain boundary. In both cases, the cracked phase boundaries contain a large number of carbides and are severely corroded, but no deformation microstructures are observed, which implies that the localized corrosion may play an important role in the IASCC along the phase boundary. In addition, δ-ferrite can retard the IASCC crack propagation along the grain boundary, which is probably related to the reduction of localized deformation by δ-ferrite.

Correlation Between IASCC Susceptibility and Magnetic Properties of Stainless Steel Irradiated with Xenon Ions

In order to investigate the correlation between irradiation-assisted stress corrosion cracking (IASCC) susceptibility and the magnetic properties of austenitic stainless steel after ion irradiation, the magnetic properties and phase transformation were studied using a vibrating sample magnetometer and synchrotron radiation grazing incidence X-ray diffraction (GIXRD) on an irradiated specimen up to 25 displacements per atom. GIXRD indicated that the body-centered-cubic (BCC) phase was introduced in austenitic stainless steel by ion irradiation. Then, magnetic hysteresis loop analysis was conducted to give a quantitative analysis on the BCC magnetic phase under different irradiation damage and annealing temperature conditions. Based on the fact that postirradiation annealing will eliminate IASCC susceptibility at 550°C for 1 h, a correlation between IASCC susceptibility and magnetic properties was discussed, and a corresponding IASCC evaluation method was developed.

A high-resolution characterization of irradiation-assisted stress corrosion cracking of proton-irradiated 316L stainless steel in simulated pressurized water reactor primary water

Irradiation-assisted stress corrosion cracks (IASCCs) from a proton-irradiated 316 L stainless steel (2.5 dpa) after constant load test in simulated PWR primary water were analyzed. The irradiation damages are consistent with those from neutron irradiations at similar dose. The IASCC cracks show microstructural and microchemical features similar to those reported from neutron-irradiated samples, confirming that proton is effective in emulating neutron for IASCC study. Si segregated at grain boundary gets oxidized preferentially beyond the crack tip and then dissolves into solution, resulting in enhanced porosity within the intergranular oxide. The Si effect is an important factor behind IASCC in reducing environment.

Hydrogen trapping by irradiation-induced defects in 316 Lstainless steel: A combined experimental and modeling study

The irradiation-induced defects in stainless steel internal components of pressurized water reactors combined with hydrogen uptake during the oxidation process could be a key parameter in the mechanism for Irradiation-Assisted Stress Corrosion Cracking (IASCC). A heat-treated 316 L SS containing a low amount of defects was Fe3+ ions-implanted; irradiation-induced defects types and depth distributions were characterized by Transmission electron microscopy (TEM). Deuterium was then inserted in the specimens either by cathodic charging or by sample exposure to deuterated primary water. Secondary Ions Mass Spectrometry – SIMS – permitted to access the deuterium distribution at the implanted stainless steel surface. A finite difference numerical solver accounting for hydrogen diffusion/trapping coupling was used to simulate the hydrogen transport in the implanted material, taken into consideration the specific heterogeneous character of the irradiation-induced defects distribution in matter. Taking as input data the experimental defects distribution associated with Frank loops or voids, the main trapping sites for hydrogen were assigned to voids, not Frank loops. Such numerical approach was able to deal accurately with the problem of hydrogen transport in a heterogeneous material as well as to differentiate two potential trap sites contributions to the experimental deuterium distribution. In addition, according to results obtained after primary water exposure, trapping at voids was still effective at 320 °C, signature of a high binding energy of hydrogen in voids.

Stress Corrosion Cracking in Austenitic Stainless Steels in Reactor Primary Water–High Purity Oxidizing and in Supercritical Conditions

Abstract Austenitic stainless steels (SS) commonly used in the primary circuit of light water reactors (LWRs) have excellent corrosion resistance in demineralized (DM) water at high temperature (290°C–330°C) and pressure (7.4 MPa for boiling water reactor [BWR], 16 MPa for pressurized water reactor [PWR]). Radiolysis of primary DM water in BWR forms 200–300 ppb of oxidizing species (normal water chemistry [NWC]), which elevates the electrochemical potential of the SS from −300 to −600 mVSHE (mV with standard hydrogen electrode) when dissolved oxygen is 10–20 ppb to +100 to +200 mVSHE. Neutron irradiation of SS further induces metallurgical and microstructural changes, which compromises corrosion resistance. Thus, radiation makes BWR (NWC) environment hostile, causing extensive irradiation-assisted stress corrosion cracking (same as intergranular stress corrosion cracking [IGSCC]) in austenitic SS. IGSCC occurs in both sensitized (grain boundary chromium depletion) and nonsensitized conditions (strain-hardened region in base metal immediately adjacent to the weld fusion zone) and is a generic problem. IGSCC in BWR can be mitigated by hydrogen addition in water (hydrogen water chemistry [HWC]). BWR-HWC has limitations and is ineffective where boiling occurs. IGSCC of SS in PWR is limited because hydrogen addition in primary water suppresses radiolysis though the formation of aggressive environment in low-flow occluded regions can cause IGSCC in SS. Increasing demand for economic power has led to the primary environment in nuclear reactors to become hostile. Conceptual supercritical water reactor (SCWR) will use DM water at 500°C–600°C and 25 MPa, which is extreme for conventional LWR materials. IGSCC in austenitic SS and nickel-based alloys occurs in both oxidizing and reducing SCWR conditions, which can be further exacerbated by radiation. This article reviews how benign primary DM water becomes hostile in LWRs and extreme in SCWR conditions, causing SCC to be a generic problem in austenitic alloys.

Compositionally graded specimen made by laser additive manufacturing as a high-throughput method to study radiation damages and irradiation-assisted stress corrosion cracking

This study demonstrates the feasibility of using compositionally gradient specimens, fabricated by laser additive manufacturing (AM) and post-AM thermo-mechanical treatment, to accelerate alloy synthesis, radiation experiment, and the assessment of irradiation properties in light water reactor environments. The effects of minor Hafnium (Hf) doping in austenitic 316L stainless steel (SS) was selected as the topic of interest. By comparing to the data in literature, we confirmed that the compositionally graded specimen produces the same trend of void swelling, dislocation loops, radiation-induced segregation (RIS), radiation hardening as the wrought specimen produced by cast/forging process. Hf suppressed most radiation damages through strong interaction with point defects. The work also demonstrates the use of compositionally gradient specimens to study the irradiation-assisted stress corrosion cracking (IASCC) susceptibility of Hf-modified SS. While the suppression of radiation hardening and RIS are consistent with the IASCC mitigation by Hf, we emphasize Hf can alter the intrinsic deformation behavior of 316L SS, which reduces grain-boundary strain localization. The advantages and challenges of using compositionally gradient design for high-throughput nuclear alloy development and qualification are also discussed.

The effect of potassium hydroxide primary water chemistry on the IASCC behavior of 304 stainless steel

Western Pressurized Water Reactors (PWRs) are considering the replacement from lithium hydroxide (LiOH) to potassium hydroxide (KOH) as a pH balancing agent in primary water due to the higher cost and lower availability of 7Li. This study qualified the irradiation-assisted stress corrosion cracking (IASCC) propagation compatibility of baffle plate materials in simulated PWR primary water with LiOH or KOH chemistries. Samples extracted from two pieces of Type 304 stainless steel baffle plates with different dose (9 and 15 dpa) were investigated. Water chemistries were switched between LiOH (1.3–2 ppm) and KOH (7.2 to 11.3 ppm) at constant pH during the crack growth rate (CGR) test on a sample, thus the effect of water chemistry was determined directly without other variations. Results indicate that there is no discernable effect on the CGR behavior by changing from LiOH to KOH.

원자로 내부구조물 균열개시 민감도에 미치는 영향인자 고찰

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

양성자 조사가 316 스테인리스강의 미세조직과 표면산화 특성에 미치는 영향

Austenitic 316 stainless steel was irradiated with protons accelerated by an energy of 2 MeV at 360 ℃, the various defects induced by this proton irradiation were characterized with microscopic equipment. In our observations irradiation defects such as dislocations and micro-voids were clearly revealed. The typical irradiation defects observed differed according to depth, indicating the evolution of irradiation defects follows the characteristics of radiation damage profiles that depend on depth. Surface oxidation tests were conducted under the simulated primary water conditions of a pressurized water reactor (PWR) to understand the role irradiation defects play in surface oxidation behavior and also to investigate the resultant irradiation assisted stress corrosion cracking (IASCC) susceptibility that occurs after exposure to PWR primary water. We found that Cr and Fe became depleted while Ni was enriched at the grain boundary beneath the surface oxidation layer both in the non-irradiated and proton-irradiated specimens. However, the degree of Cr/Fe depletion and Ni enrichment was much higher in the proton-irradiated sample than in the non-irradiated one owing to radiation-induced segregation and the irradiation defects. The microstructural and microchemical changes induced by proton irradiation all appear to significantly increase the susceptibility of austenitic 316 stainless steel to IASCC.

Effects of Radiation-Induced Defects on Corrosion

The next generation of nuclear reactors will expose materials to conditions that, in some cases, are even more extreme than those in current fission reactors, inevitably leading to new materials science challenges. Radiation-induced damage and corrosion are two key phenomena that must be understood both independently and synergistically, but their interactions are often convoluted. In the light water reactor community, a tremendous amount of work has been done to illuminate irradiation-corrosion effects, and similar efforts are under way for heavy liquid metal and molten salt environments. While certain effects, such as radiolysis and irradiation-assisted stress corrosion cracking, are reasonably well established, the basic science of how irradiation-induced defects in the base material and the corrosion layer influence the corrosion process still presents many unanswered questions. In this review, we summarize the work that has been done to understand these coupled extremes, highlight the complex nature of this problem, and identify key knowledge gaps.

Radiation effects related to repaired BWR core shrouds

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

IASCC of neutron irradiated 316 stainless steel to 125 dpa

The irradiated microstructure and irradiation-assisted stress corrosion cracking (IASCC) susceptibility were determined for neutron-irradiated cold-worked (CW) 316 stainless steel tested in 320 °C simulated PWR primary water with various damage levels of 46.9, 67.4, and 125.4 dpa. The size and density of dislocation loops and precipitates were found to vary little over the dose range 46.9–125.4 dpa and dislocation loops were found to control the irradiation hardening. The %IG in constant extension rate tensile (CERT) tests and the stress to initiate a crack in four-point bend tests were consistent in establishing that the IASCC susceptibility was essentially unchanged across the dpa range. GB oxidization was observed prior to crack initiation, and localized deformation and triple junctions where local stresses were high were preferred crack initiation sites. IASCC cracks initiated preferentially on grain boundaries oriented nearly normal to the applied stress. An apparent threshold stress threshold for IASCC was determined to be ~45% of the irradiated yield strength for neutron-irradiated CW 316 SS between 46.9 and 125.4 dpa.

A micromechanical analysis of intergranular stress corrosion cracking of an irradiated austenitic stainless steel

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a material degradation phenomenon affecting austenitic stainless steels used in nuclear Pressurized Water Reactors (PWR), leading to the initiation and propagation of intergranular cracks. Such phenomenon belongs to the broader class of InterGranular Stress Corrosion Cracking (IGSCC). A micromechanical analysis of IGSCC of an irradiated austenitic stainless steel is performed in this study to assess local cracking conditions. A 304L proton irradiated sample tested in PWR environment and showing intergranular cracking is investigated. Serial sectioning, Electron BackScatter Diffraction (EBSD) and a two-step misalignment procedure are performed to reconstruct the 3D microstructure over an extended volume, to assess statistically cracking criteria. A methodology is also developed to compute Grain Boundary (GB) normal orientations based on the EBSD measurements. The statistical analysis shows that cracking occurs preferentially for GB normals aligned with the mechanical loading axis, but also for low values of the Luster-Morris slip transmission parameter. Micromechanical simulations based on the reconstructed 3D microstructure, FFT-based solver and crystal plasticity constitutive equations modified to account for slip transmission at grain boundaries are finally performed. These simulations rationalize the correlation obtained experimentally into a single stress-based criterion. The actual strengths and weaknesses of such micromechanical approach are discussed.