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ABSTRACT Mechanical properties such as yield strength and hardness of irradiated stainless steels can be significantly changed by increasing irradiation dose. An understanding of the effects of mechanical properties on crack growth in neutron-irradiated stainless steels is required for identifying the irradiation assisted stress corrosion cracking () mechanism and for establishing strategies for mitigation. In this study, experimental crack growth rate () data of irradiated alloys reported by the university and the Cooperative Research () program were correlated with hardness and yield strength, as well as with microstructural and microchemical variables such as loops, precipitates, and Cr depletion and Si enrichment at grain boundaries. It was found that the increase in CGR was proportional to the increase in yield strength and hardness caused by irradiation, but not directly related to many other variables. To simulate the effect of irradiation on CGR, finite element method (FEM) analyses on yield strength increased alloys were conducted to estimate the size and shape of plastic zone at the crack tip. Yield strengths of 250, 500, 750 and 1000 MPa were used to simulate stainless steels at different irradiation doses. An 8 mm round compact tension () specimen under constant K control at 15 MPa·mwas used in the FEM simulation. The CGR in high strength irradiated alloys correlates with the decrease in plastic zone size, the decrease in tensile plastic strain and strain increment, and the increase in tensile "opening" stress. In addition, similarities and differences between cold work hardening and radiation hardening are discussed.
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- Materials > Metals & Mining > Steel (1.00)
- Energy > Power Industry > Utilities > Nuclear (0.47)
ABSTRACT: To better understand the irradiation assisted stress corrosion cracking (IASCC) mechanism and to establish strategies for IASCC mitigation, it is essential to acquire accurate crack growth data on neutron-irradiated alloys in the relevant environment. In this study, a crack growth rate (CGR) test on a neutron-irradiated stainless steel alloy was conducted under constant K condition in BWR NWC, BWR HWC, and PWR primary water at 288°C and 320°C. The specimen was an 8 mm round compact tension (RCT) sample of high purity 316L SS containing Hf (HP316L+Hf), irradiated to 9.6 dpa in the BOR-60 fast reactor. The CGR test was conducted under load and environmental control and for duration of 3127 hours. The CGR was characterized by real time recording of DCPD and the post-test measurement of actual crack length on the fracture surface. The test results indicated that the dependence of CGR on K was 5.1 in the K range of 13-22 MPa.m1/2 in BWR NWC. CGRs in BWR HWC are more than one order of magnitude lower than those in BWR NWC at the equivalent K level. INTRODUCTION In recent years, CGR data on neutron irradiated alloys have been reported in international cooperative programs. 6, 8-15 In the CIR program, 9-14 twenty CGR tests on neutron-irradiated commercial alloys and tailored alloys were reported by Studsvik, NRI, and SCK-CEN. CGR tests conducted at Studsvik were on commercial alloys. 9, 10 High purity tailored solute addition alloys were tested at NRI 14 and SCK-CEN.12 These alloys include alloy ES (a reference, high purity alloy of composition Fe-18Cr-12Ni-1Mn), and alloys HS (+Si), KS (+Ni), PS (+Hf) tested at NRI, and FS (-C), GS (+Mo), LS (+Cr+Ni) tested at SCK-CEN. There was no consistent effect of solute addition on CGR among ES, HS, and PS in BWR water.
- Materials > Metals & Mining > Steel (0.91)
- Energy > Power Industry > Utilities > Nuclear (0.68)
Repassivation Behavior And Stress Corrosion Cracking Susceptibility Of Stainless Steels Containing Silicon
Chou, Peter H. (Electric Power Research Institute,GE Global Research 1 Research Circle) | Andresen, Peter L. (GE Global Research ) | Pollick, Michael L. (GE Global Research ) | Rebak, Raul B. (GE Global Research )
ABSTRACT Stress corrosion cracking (SCC) is a pervasive failure mode for austenitic stainless steels core internals components in nuclear light water reactors. Some cases of SCC may be exacerbated by irradiation. Irradiation may change in the dislocation distribution and increase the yield stress (hardness) of the materials. Irradiation may also alter the local chemistry of the austenitic alloys, for example causing chromium depletion and silicon enrichment at the grain boundaries. The objective of the present work was to perform laboratory tests in order to better understand the role of Si on the microstructure, electrochemical behavior and susceptibility to SCC of austenitic stainless steels. Little or no effect was found on the effect of Si on the electrochemical behavior in high temperature water of type 304 SS containing from less than 1% Si up to 5% Si in the bulk. Similarly, current SCC crack growth rate results are not conclusive regarding a consistent effect of the bulk concentration of Si on the SCC resistance of the stainless steels. INTRODUCTION Austenitic stainless steels (SS) core internals components are susceptible to irradiation assisted stress corrosion cracking (IASCC) during service in nuclear power plants light water reactors. One of the effects of irradiation is the hardening of the SS due to modifications in the dislocation distribution in the alloy. Irradiation also alters the local chemistry of these austenitic alloys, for example in the vicinity of grain boundaries by a mechanism of radiation induced segregation (RIS). The segregation or depletion phenomena at near grain boundaries may enhance the susceptibility of these alloys to stress corrosion cracking. As the amount of damage by radiation increases, austenitic stainless steels become increasingly vulnerable to SCC. For example, it has been suggested that thermally non-sensitized irradiated SS becomes depleted in chromium (Cr) at the grain boundaries and subsequently becomes more susceptible to SCC., It is also apparent that the susceptibility of SS to SCC is highly dependent on small chemistry and process variations that are typical from different heats of the same material (heat-to-heat variability). It has been many times claimed that the segregation to the grain boundaries of elements such as carbon (C), sulfur (S), phosphorus (P), oxygen (O), nitrogen (N) and silicon (Si) are important factors on the resistance of SS to SCC. However, the role of the segregation of these minor alloying elements on SCC is still poorly understood. Under irradiation, Si may enrich at the grain boundary up to ten times its bulk composition. The role of Si and its grain boundary segregation on the IASCC remains contradictory. Jacobs et al. did not find any correlation between the amount of Si segregated at the grain boundaries in Type 348 SS and the percentage of intergranular stress corrosion cracking on the surface using constant extension rate tests in high purity water at 288°C containing 32 ppm of dissolved oxygen.
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- Materials > Metals & Mining > Steel (1.00)
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- Energy > Oil & Gas > Upstream (1.00)
INTRODUCTION ABSTRACT Numerous mechanisms of crack advance have be-en proposed for iron- and nickel-base alloys in high temperature water, with most focusing on a slip / film rupture / oxidation or a direct hydrogen mechanism of crack advance. Numerous roles for hydrogen have been proposed, including effects on dislocation mobility, on the cohesive strength of metal bonds, on void I bubble I blister formation, etc.; some of these roles could influence other mechanisms of crack advance. The objective of this paper is to review the evidence of hydrogen effects in environmental cracking in hot water. While additional critical tests are needed to more precisely define its role, present evidence provides only weak support that hydrogen plays a predominant role in environmental cracking in hot water. Environmental cracking in hot water continues to receive strong attention because of the large technological importance of hot water systems, especially in the energy industries. Specific applications include various components in fossil boilers, boiling water reactors, pressurizing water reactors, steam turbines, deaerators, etc. Most of the structural components are made of iron- and nickel-base alloys, which suffer from sub-critical crack advance during exposure to high temperature water ranging from <100ºC to >360ºC. Because of its ubiquitous nature, ability to diffuse in metals, and many interactions in metals, hydrogen is rightfully viewed as a likely contributor to many forms of environmental degradation [1-4]. Many of the well established roles of hydrogen, such as hydrogen induced localized plasticity and reduction in the cohesive strength of metal bonds, become much less compelling as the test temperature is increased above about 150ºC. In turn, effects of hydrogen that can be important at higher temperature, such as hydride or methane gas bubble formation, am unlikely contributor across the range of iron- and nickel-base alloys and temperatures where environmental cracking is commonly observed in hot water. An alternative view of environmentally assisted cracking in hot water is that it occurs by a mechanism that does not rely directly on hydrogen. Among these, the slip / film rupture / oxidation mechanism is the most prominent [5-7]. This mechanism attributes crack advance directly to corrosion processes at the crack tip, with various environmental, metallurgical, and mechanical variables affecting the localized “crack tip system”. These effects can be divided into mechanical effects that control the crack tip strain rate and thereby the Periodicity of oxide rupture (e.g., loading, creep rate, yield strength, oxide ductility) and chemical effects that determine the repassivation response at the crack tip (e.g., crack tip microstructure and microchemistry, and crack tip chemistry, which is influenced by corrosion potential, impurities, flow rate, etc.). Various observations will be reviewed and discussed that reflect on the role for hydrogen as the direct, causal mechanism of crack advance, or as a significant contributor to other crack advance mechanisms, e.g., via enhancement of dislocation mobility and associated creep. ROLES OF HYDROGEN Many conceptual mechanisms by which hydrogen can affect behavior of metals in general and environmental cracking in specific have been advanced [14].
ABSTRACT The effect of zinc (Zn) and copper (Cu) additions on the catalytic behavior of noble metal alloyed 304 stainless steel (SS) in 288C water under stoichiometric excess hydrogen was studied. It was observed that an increase in the Zn or Cu content of the water increased the electrochemical corrosion potential (ECP) of noble metal alloyed 304 SS by = 30 to 50 mV and decreased the recombination efficiency of oxygen (O2) and hydrogen (H2) by =10%. The change in the ECP and recombination rate was correlated with incorporation of zinc and copper in the oxide film, which, by covering catalytic sites, would alter the redox reaction rate. INTRODUCTION The high temperature water in boiling water reactors (BWRs) is highly oxidizing due to radiolytically generated dissolved O2 and hydrogen peroxide (H2O2) (100 to 300 ppb) and dissolved H2 (10 to 20 ppb). In this environment, intergranular stress corrosion cracking (IGSCC) of sensitized SS components in BWRs has been a major concern [1, 2]. The IGSCC susceptibility of reactor structural materials in BWRs is known to be affected by the ECP which is controlled by the content of the total oxidant such as O2 and H2O2 Data from laboratories and reactors have shown that IGSCC can be prevented by reducing the oxidant concentration and thereby lowering the ECP below -230 mV vs. the saturated hydrogen electrode (SHE) [4-6]. H2 is being added to the feedwater of BWR to mitigate the IGSCC by reducing the dissolved oxidant concentration. This process is referred to as hydrogen water chemistry (HWC). Large amounts of H2 addition are normally required to sufficiently lower the dissolved O2 and H2O2 concentrations so that the IGSCC protection potential (-230 Mvshe) is attained. Although IGSCC can be sufficiently suppressed by the HWC process, it is not an optimum process from an operating viewpoint. The disadvantages of HWC include high H2 cost, higher Co buildup rate, increased N release to the turbine, etc. The IGSCC protection potential is also difficult to achieve in the highly oxidizing and high water flow regions. Zinc oxide (ZnO) has been added into the BWR feedwater because of the benefical implication in reducing buildup of radiactive species in the oxide film [7, 8]. It was reported that Zn addition to high temperature water altered the oxide structure and enhanced a thin, less defect corrosion film on 304 SS [9], decreased the growth rate of IGSCC of SS and nickel-base alloys [10] and increased the oxide rupture strain of 304 SS [11]. Also, different levels of Cu had been detected in plants which mainly came from the corrosion of brass condenser tube. Recently, noble metal technology (NMT) has been developed to achieve the thermodynamically lowest possible ECP (-500 mVshe)and lowest crack initiation/growth rates at much lower H2 addition rates and with minimal negative impact on BWR operation [12-16]. This approach invloves improving the catalytic recombination of O2 and H2O2 with H2 to from H2O on metal surfaces.
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