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Collaborating Authors
Andresen, Peter L.
ABSTRACT The influence of material and applied stress on stress corrosion cracking (SCC) were evaluated to 18,500 hours on candidate waste package materials for the Yucca Mountain Project. Time-to-failure experiments were being performed on smooth bar tensile specimens in a hot, concentrated, mixed-salt solution chosen to simulate concentrated Yucca Mountain water. The effects of applied stress, welding, surface finish, shot peening, cold work, crevicing, and aging treatment are investigated for Alloy 22 (UNS N06022). Aging treatments were designed to produce topologically close-packed phases (TCP) and long-range ordering (LRO) as worse-case scenarios for possible microstructures in Alloy 22. Titanium Grade 7 and type 316NG stainless steel were also included in the matrix, as they were also being considered for drip shield and waste package components, respectively. Sensitized type 304 stainless steel was evaluated to provide benchmark data. INTRODUCTION General corrosion, localized corrosion (pitting and crevice corrosion) and stress corrosion cracking are the most likely degradation modes for nuclear waste package structural materials. Since the waste package will always be hotter than its surrounding environment, the solution composition on the waste package surface will concentrate as the waste package initially cools to the temperature of the maximum boiling point elevation. As the Yucca Mountain water drips or splashes onto the drip shield and later the waste package and concentrates, the pH of typical carbonate-rich seepage water is expected to rise to at least 10 and perhaps higher. This research study is designed to use a statistical approach in determining the life and reliability of proposed waste package materials potentially susceptible to stress corrosion cracking under conditions that are both relevant and likely to promote stress corrosion cracking, i.e., fairly high concentrations at fairly high temperatures. The experimental results, a continuation of work reported in reference [1] provide a means of making probabilistic comparisons between material-conditions. These results are strengthened with crack growth data obtained in a parallel study [2]. Other programs are addressing the general and localized corrosion in this and related environments [3-6]. EXPERIMENTAL PROCEDURE On-going stress corrosion crack initiation experiments to obtain time-to-failure data are being performed on smooth tensile specimens of various materials (Table 1). Each specimen is individually and actively loaded by the internal pressure of a large Keno autoclave. A schematic of the Keno autoclave cross-section is shown in Figure 1. The type 347 stainless steel autoclave has a volume of 68 liters, which is filled with mixed salt solution. The composition of the mixed salt solution used in this study was chosen to simulate concentrated Yucca Mountain water (Table 2). Three type 304 stainless steel manifolds are loaded into the Keno autoclave, where each manifold can support fifty type 304 stainless steel specimen module assemblies (a total of 150 specimens). A schematic of a module assembly is shown in Figure 2. The load on each specimen is created by the pressure differential across a sliding seal on a piston connected to the specimen, where internal pressure of the autoclave is on one side and atmospheric pressure is on the other side of the manifold. A pressuring gas creates the internal pressure in the autoclave due to the lack of substantial vapor pressure at 105?125oC. On specimen failure, the piston and specimen cause a numbered indicator ball to be ejected into the manifold. Time to failure is registered when the indicator ball runs down the manifo
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ABSTRACT The stress corrosion crack growth rate response was evaluated on as-received, cold worked and aged Alloy 22 (UNS N06022) in 110 C, aerated, concentrated, high pH groundwater environments. Long term crack growth rate data showed that Alloy 22 exhibited stable growth rates under gentle cyclic loading, but was prone to crack arrest at fully static loading, unlike observations in prior work on Ti Grade 7. No effect of Pb additions was observed. INTRODUCTION Alloy 22 (UNS N06022) is the current reference material for the outer barrier in the high level nuclear waste package for the Yucca Mountain Project. The waste package is an essential element of the engineered barrier system, and the ability to provide very long waste package lifetimes that can be predicted with confidence is a central factor in the release rate of radionuclides to the accessible environment. Stress corrosion cracking (SCC), along with general and localized corrosion, are the most likely degradation modes for the waste package materials. While aggressive environments that give rise to pitting and crevice corrosion may also induce stress corrosion cracking, it has proven incorrect to assume that because the material is highly resistant to localized corrosion that it is also stress corrosion resistant. Increasingly careful SCC studies reveal that many materials once thought to be immune do exhibit stress corrosion crack growth under constant stress intensity factor conditions [1,2]. Prior studies showed this to be the case for titanium Grade 7 tested in these environments [3], where well-behaved, sustained SCC growth was observed over thousands of hours (Figure 1). Further, Alloy 22 (UNS N06022), although extremely resistant to localized corrosion over a broad range of potential concentrated brine environments [4], is not fully immune to SCC growth. However, the extremely low measured crack growth rates described in this paper indicate the alloy is highly resistant to crack growth under constant stress intensity conditions. Since no evidence of SCC initiation has been observed on smooth and defected specimens over a broad range of metallurgical conditions and applied stresses as determined in parallel testing programs [5,6] and considering the highly static loading conditions representative of the waste package, it is exceedingly unlikely that Alloy 22 SCC will occur under Yucca Mountain Repository environmental and loading conditions. Nevertheless, the Project approach is to preclude outer surface tensile stresses by utilizing an effective stress mitigation process [7]. Demonstrating and predicting corrosion damage and associated waste package lifetimes depends primarily on characterizing the local environment that forms on the waste package. This is particularly important at higher temperatures (above 75 C), where the heat flux through the waste package is higher, the environments more concentrated, and the material susceptibility to corrosion degradation highest. Because of the radioactive decay heat generated within the waste packages, there is a resulting heat flux across each waste package and adjacent drift wall that results in the waste package always being somewhat hotter than its surrounding environment.. Since it is reasonable to assume that water reaches the emplacement drift, and that the higher surface area of the tunnel walls controls the air temperature and maintains relative humidity near 100%, any liquid that forms on the waste package must concentrate sufficiently to account for the temperature differential between the emplacement drift wall and the waste package. Whether from dripping / splashing ground water, contaminants from handling, or rock du
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Development and the Use of a Lee James Surface Cracked Specimen for Evaluating Chemistry and Flow Rate Effects in Realistic Cracks
Andresen, Peter L. (GE Global Research Center) | Ayhan, Ali (GE Global Research Center) | Catlin, Gregory (GE Global Research Center) | Catlin, Bill (GE Global Research Center) | Miller, William D. (GE Nuclear Energy)
ABSTRACT Modifications and finite element analyses have been performed to adapt a surface cracked specimen developed by Lee James. The specimen design produces very tight, linear cracks that are approximately semicircular and which closely simulate the nature of cracks that form in plant components. This is particularly important for studying flow rate effects. INTRODUCTION Stress corrosion cracking (SCC) has occurred in many structural materials used in boiling water reactors (BWRs) and pressurized water reactors (PWRs). This has spawned extensive evaluation of the stress corrosion crack growth behavior of various grades of stainless steel, alloy 600, alloy 182 weld metal, and low alloy and carbon steels under unirradiated and irradiated conditions [1-10]. From the earliest studies it was clear that corrosion potential had a very strong influence on SCC in high temperature water (Figure 1), and methods to decrease the corrosion potential were pursued for SCC mitigation. Simple injection of H2 into the feed water of BWRs provides some reduction in corrosion potential, but high H2 levels are often required, a large increase in 16N turbine shine often results at high H2, and an adequate reduction in the corrosion potential in the core is not always achievable. For these reasons and others, noble metal technology was developed. It provides a unique opportunity for achieving the thermodynamically lowest possible corrosion potential and therefore the lowest possible SCC growth rates with minimal negative impact on BWR operation [2,7,11-17]. Noble metals are electrocatalysts that efficiently recombine O2 and H2O2 with H2 on the metal surface. Once a near-stoichiometric concentration of H2 is present for the formation of water (2H2 + O2 ?¨ 2H2O), the corrosion potential decreases to its thermodynamic minimum value of ¡Ö ?0.52 Vshe (in 288 °C pure water containing ¡Ö 100 ppb H2) . This nominally occurs at a 2:1 H:O molar ratio, which corresponds to a 1:8 H:O weight ratio, so that excess H2 exists if its concentration is greater than one-eighth of the O2 value (e.g., in ppb). However, slightly sub-stoichiometric H2 (e.g., 1:10 to 1:12 H:O) is sufficient because the diffusivity of H2 in the stagnant liquid boundary layer is higher than O2 or H2O2. A variety of local and system-wide coating techniques have been developed for creating catalytically active surfaces on structural materials using noble metals [11-17]. Layers of very high catalyst concentration have been produced by electro- and electro-less plating, vapor deposition, sputtering, etc. of pure or mixed noble metals. A system-wide approach which relies on electroless reduction of very dilute noble metal compounds added to the BWR coolant for a limited time period, has been termed in-situ noble metal chemical addition [11- 17], and shows great promise as a technique to provide a noble metal coating on all wetted components. This catalytic effect has been demonstrated for a broad range of iron, nickel, and cobalt-based noble metal alloys, materials coated with individual or mixtures of noble metals [11-17]. In addition to a catalytic surface and stoichiometric excess H2, the fundamental criterion for the success of noble metal technology is that the catalysis extends at least as deep into a crack as does O2 (or other oxidants). This produces a low corrosion potential everywhere in the crack (Figure 2a), thereby eliminating the gradient in corrosion potential that produces an aggressive crack chemistry and accelerates SCC growth rates (Figure 1). However, even if low potential conditions exist everywhere, a sufficiently aggressive chemistry will still accelerate SCC (Figure 2b). To evaluate the loss of
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Stress Corrosion Crack Growth Rate Response of AH & HTH Alloy X750 In High Temperature Water
Andresen, Peter L. (GE Corp. R&D Center) | Emigh, Paul W. (GE Corp. R&D Center) | Horn*, Ron M. (GE Nuclear Energy) | Gordon, Gerald (General Electric Nuclear Energy) | Young, Lisa (GE Global Research Center)
ABSTRACT The stress corrosion crack growth rate response of alloy X750 was studied in high temperature water as a function of loading, corrosion potential, water purity, and heat treatment condition (AH and HTH). In addition to providing more quantitative data on the behavior of alloy X750, the response of this higher strength material is interesting to compare with similar materials whose yield strength is varied by cold work or irradiation. INTRODUCTION Stress corrosion cracking (SCC) has occurred in many materials used in boiling water reactors (BWRs) and pressurized water reactors (PWRs). There has been extensive evaluation of the crack growth behavior of various grades of stainless steel, alloy 600, alloy 182 weld metal, and low alloy and carbon steels [1-4]. However, there is only limited stress corrosion crack growth rate data on alloy X750 in high temperature, high purity water [1 ]. While high purity, BWR water environments were the focus of the work, it should be recognized that there is strong applicability to PWR environments. Because the crack tip is deaerated and at low potential in all cases (with the increasing adoption of NobleChem TM in BWRs, even the surface corrosion potential is low), the environmental conditions under which crack advance occurs in BWR and PWR primary systems are very similar [2-7]. Thus, BWRs and PWRs differ primarily in: coolant additives that shift the pH at temperature from 5.6 to --7.0; H2 fugacity (--50 vs 3000 ppb H2); and temperature (since most structural materials in a BWR are exposed to 274 °C water, the PWR primary is up to 50 °C hotter, 65 °C in the PWR pressurizer). The factor with the most pronounced effect on SCC is temperature. The evaluation of alloy X750 is also applicable to an improved understanding of the role of yield strength in SCC, which is important because a broad range of yield strengths exist in structural materials because of cold work, weld shrinkage strains, irradiation hardening, precipitation hardening, etc. Prior work on the effects of yield strength by cold work [8-12], weld shrinkage strains [8-12] and irradiation [6,13] in austenitic 304/304L/316L stainless steels has shown that 20% reduction in thickness (resulting in about 80 ksi (552 MPa) yield strength) can produce large enhancements in crack growth rate (Figure * 1). At high potential, these rates are equivalent to those observed on sensitized stainless steel, while at low potential the rates are roughly an order of magnitude higher (Figures 1 - 5). Essentially identical effects of yield strength were observed on alloy 600 (Figure 6). The crack length vs. time response is generally very well behaved at both high potential and low potential (Figures 2 and 3), even at low stress intensity factor (Figure 4) and/or with weld shrinkage strains (Figure 5). Stress corrosion cracking is intergranular in nature despite the absence of Cr depletion or other detectable grain boundary inhomogeneity (Figure 6). Because the effect of chromium depletion (and presumably some other forms of chemical inhomogeneities at the grain boundary) is diminished at low corrosion potential, a comparison was made between the effect of yield strength by irradiation and by cold work [13]. While exact one-to-one comparison data are not available, multiple, reasonably close comparisons could be made to support the similarity in response between cold worked and irradiated stainless steel of similar yield strength. One side benefit of this study is to compare the response of an alloy whose yield strength is elevated by precipitation hardening. Follow up studies may evaluate the same alloy X750 materials whose yield strength is enhanced only by cold work. EXPERIMENTAL
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ABSTRACT This paper provides an update on research addressing the effects of material condition and applied stress on stress corrosion cracking (SCC) in waste package and drip shield materials for the Yucca Mountain Project. Time-to-failure experiments are being performed on smooth bar tensile specimens in a hot, concentrated, mixed-salt solution chosen to simulate concentrated Yucca Mountain water. The effects of applied stress, welding, surface finish, shot peening, cold work, crevicing, and aging treatment are being investigated for Alloy 22 (UNS N06022). Aging treatments were designed to produce topologically close-packed phases (TCP) and long-range ordering (LRO) and are under investigation as worse-case scenarios for possible microstructures in Alloy 22 (UNS N06022). Titanium Grade 7 and 316NG stainless steel are included in the matrix, as they are identified for drip shield and waste package components, respectively. Sensitized 304SS specimens are included in the test matrix to provide benchmark data. This research complements high-resolution crack-growth-rate experiments currently being performed in a parallel research project. INTRODUCTION General corrosion, localized corrosion, and stress corrosion cracking represent the most likely degradation modes for nuclear waste package structural materials. Since the waste package will always be hotter than its surrounding environment, the solution composition on the waste package surface will concentrate as the waste package initially cools to the temperature of the maximum boiling point elevation. As the Yucca Mountain water drips or splashes onto the drip shield and later onto the waste package where it concentrates, the pH of typical carbonate-rich seepage water is expected to rise to at least 10 and perhaps higher. This research study is designed to use a statistical approach in determining the life and reliability of proposed waste package materials potentially susceptible to stress corrosion cracking under environmental conditions that are both relevant and likely to promote stress corrosion cracking, i.e., fairly high solution concentrations at fairly high temperatures. The experimental results will provide a means of making probabilistic comparisons between material-conditions. These data will be strengthened with laboratory crack growth data obtained in a parallel study [1,2]. Other programs are addressing the general and localized corrosion in this and related environments [3-6]. EXPERIMENTAL PROCEDURE Time-to-failure experiments are being performed on smooth tensile specimens that are individually loaded by the internal pressure of a large Keno autoclave. A schematic of the Keno autoclave cross-section is shown in Figure 1. The 347SS autoclave has a volume of 68 liters, which is filled with mixed salt solution. The composition of the mixed salt solution used in this study was chosen to simulate concentrated Yucca Mountain water (Table 1). Three 304SS manifolds are loaded into the Keno autoclave, where each manifold can support fifty 304SS specimen module assemblies (for a total of 150 specimens). A schematic of a module assembly is shown in Figure 2. The load on each specimen is created by the pressure differential across a sliding seal on a piston connected to the specimen, where internal pressure of the autoclave is on one side and atmospheric pressure is on the other side of the manifold. A pressuring gas is used to create the internal pressure in the autoclave due to the lack of substantial vapor pressure at 105-125oC. On specimen failure, the piston and specimen cause a numbered indicator ball (i.e., Keno) to be ejected into the manifold. The indicator ball runs down the manifold into a track, past a sensor (which re
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ABSTRACT Stress corrosion crack growth rate data have been obtained on as-received, cold worked and aged Alloy 22 (UNS N06022) in 110 C, aerated, concentrated, high pH groundwater environments. The long term crack growth rate behavior under very stable conditions was monitored using reversing dc potential drop. Alloy 22 exhibited stable growth rates under "gentle" cyclic loading, but was prone to crack arrest at fully static loading. Reference to related data on Ti Gr.7 is also made. INTRODUCTION Alloy 22 (UNS N06022) is the current reference material for the outer barrier of the high level nuclear waste package for the Yucca Mountain Project. The waste package is an essential element of the engineered barrier system, and the ability to provide very long waste package lifetimes that can be predicted with confidence is a central factor in the release rate of radionuclides from the mountain. General and localized corrosion, and stress corrosion cracking (SCC) represent the most likely degradation modes for the materials comprising the waste package. While aggressive environments that give rise to pitting and crevice corrosion may also induce stress corrosion cracking, it has proven incorrect to assume that because the material is highly resistant to localized corrosion that it is also highly stress corrosion resistant. Indeed, increasingly careful SCC studies reveal that many materials once thought to be immune do exhibit stress corrosion crack growth under constant stress intensity factor conditions [1,2]. Prior studies showed this to be the case for titanium Grade 7 tested in these environments [3], where well-behaved, sustained SCC growth was observed over thousands of hours (Figure 1). One key to demonstrating and predicting waste package lifetimes lies in characterizing the local environment that forms on the waste package. This is particularly important for higher temperatures (above -- 75 °C), where the heat flux through the waste package is higher, the environments more concentrated, and the material susceptibility to corrosion degradation highest. The waste package is always hotter than its surrounding environment, probably by several degrees C at > 70 °C. Since it is reasonable to assume that water reaches the emplacement drift, and that the higher surface area of the tunnel walls controls the air temperature and maintains relative humidity near 100%, any liquid that forms on the waste package must concentrate sufficiently to account for the temperature differential (e.g., 2 - 5 °C) between the emplacement drift wall and the waste package. Whether from dripping / splashing ground water, contaminants from handling, or rock dust and atmospheric aerosols (during construction / ventilation), it is hard to preclude the prospect of aqueous films forming on the waste package. The most likely scenario relates to evaporative concentration of seepage waters and/or deliquescence of soluble species in surface dust deposits on the titanium grade 7 drip shield and/or the Alloy 22 waste package that can lead to solution concentrations of at least several molar on the metal surfaces. The concentration of any aqueous phase will decrease with time / temperature, although even after >> 10,000 years, a waste package temperature of 40 or 50 °C is expected, and this brings with it an aqueous phase of about one molar solution. As the mobile carbonate dominated water from Yucca Mountain concentrates, its mixed ion character remains intact, and its pH rises to at least 9 or 10. This is likely to be more benign than a hot, concentrated, acidified chloride that was envisioned early in the Project. The general and localized corrosion in these and related environments are being studied in o
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ABSTRACT Enhancing the electrocatalytic properties of wetted surfaces is a very effective means of mitigating SCC in oxidizing environments such as in BWRs, provided there is a stoichiometric excess of reductants over oxidants. The mechanisms and criteria for effective SCC mitigation are summarized, with particular focus on the critical location for the catalyst. stress corrosion cracking, high temperature water, catalysis, noble metals, crack growth rate, corrosion potential, structural materials, boiling water reactors. INTRODUCTION Stress corrosion cracking (SCC) has occurred in many structural materials used in boiling water reactors (BWRs). There has been extensive evaluation of the crack growth behavior of various grades of stainless steel, alloy 600, alloy 182 weld metal, and low alloy and carbon steels under unirradiated and irradiated conditions [1-10]. From the earliest studies it was clear that corrosion potential had a very strong influence on SCC in high temperature water (Figure 1), and methods to decrease the corrosion potential were pursued for SCC mitigation. Noble metal technology provides a unique opportunity for achieving the thermodynamically lowest possible corrosion potential and therefore the lowest possible SCC growth rates with minimal negative impact on BWR operation [2,7,11-17]. This paper summarizes the mechanism by which noble metals act in high temperature water in terms of electrochemical kinetics. The kinetic requirement in typical gas phase catalysis where reaction rates is very high compared to BWR systems where surface reaction rates are much lower due to the presence of a liquid phase and lower concentrations of reactants (i.e., O2, H202, and H2). Thus, dilute noble metal alloys (e.g., stainless steel containing 0.1% Pd, Pt...) and/or low surface loadings also behave catalytically. Noble metals are electrocatalysts that efficiently recombine O2 and H202 with H2 on the metal surface by providing surface sites on which these species can dissociatively adsorb and readily undergo electron exchange reactions; the undissociated molecules are relatively stable when homogeneously dispersed at BWR temperatures. Once a near-stoichiometric concentration of H2 is present for the formation of water (2H2 + 02 ~ 2H20), the corrosion potential decreases to its thermodynamic minimum value of---0.52 gshe (in 288 °C pure water containing --0.01 -0.1 atm. H2). This nominally occurs at a 2:1 H:O molar ratio, which corresponds to a 1:8 H:O weight ratio, so that "excess H2 exists if its concentration is greater than one-eighth of the 02 value (e.g., in ppb). However, slightly sub- stoichiometric H2 (e.g., 1:10 to 1:12 H:O) is sufficient because the diffusivity of H2 in the stagnant liquid boundary layer is higher than 02 or H202. A variety of local and system-wide coating techniques have been developed for creating catalytically active surfaces on structural materials using noble metals [11-17]. Layers of very high catalyst concentration have been produced by electro- and electro-less plating, vapor deposition, sputtering, etc. of pure or mixed noble metals. A system-wide approach which relies on electroless reduction of very dilute noble metal compounds added to the BWR coolant for a limited time period, has been termed "in-sire noble metal chemical addition" [11-17], and shows great promise as a technique to provide a noble metal coating on all wetted components. This catalytic effect has been demonstrated for a broad range of iron, nickel, and cobalt-based noble metal alloys, materials coated with individual or mixtures of noble metals [ 11-17]. The purpose of this paper is to discuss the mechanistic role of corrosion potential in influencing
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Passivity of Nuclear Waste Canister Candidate Materials in Mixed-Salt Environments
Kim, Young- Jin (General Electric Corp. R&D) | Andresen, Peter L. (General Electric Corp. R&D) | Martiniano, Paul (General Electric Corp. R&D) | Chera, John (General Electric Corp. R&D) | Larsen, Michael (General Electric Corp. R&D) | Gordon, Gerald M. (Framatome ANP/Bechtel ASIC Company)
INTRODUCTION ABSTRACT The passivity behavior of Alloy 22 and Grade 7 titanium has been studied at 95°C in a high pH salt environment characteristic of concentrated Yucca Mountain groundwater. Measurements of corrosion potential (CP) versus time, potentiostatic polarization (PP) and cyclic potentiodynamic polarization (CPP) behavior were conducted to evaluate the passivity of these alloys. The characterization of passive films was also analyzed by x-ray photoelectron spectroscopy (XPS) and transmission electron microscopy (TEM) to obtain the chemical composition and cross-sectional view of the metal, interface, and oxide layers. It was observed that the oxide layer responsible for passivity of Alloy 22 consisted of chromium oxide (Cr203) containing Ni. The surface analysis showed that the passive films formed on Alloy 22 at high anodic potentials (> 0 mV vs. SCE) contained more Mo and W than ones formed at lower anodic potentials (<0 mV vs. SCE). However, no visual evidence of localized corrosion on Alloy 22 after potentiostatic polarization measurements was observed. This program is designed to examine the oxide characteristics formed on Alloy 22 and Ti-grade 7 as disposal-canister and drip-shield corrosion resistant materials, respectively for the Yucca Mountain Project [ 1]. The waste package and drip shield are essential elements of the engineered barrier system, and the ability to provide very long waste package lifetimes that can be predicted with confidence is a central factor in the calculated release rate of radionucleides from the mountain. General and localized corrosion, and stress corrosion cracking represent the most likely degradation modes for the corrosion-resistant materials comprising the waste package and drip shield [2- 5]. One key to demonstrating and predicting waste package lifetimes lies in characterizing the local environment that forms on the waste package. This is particularly important for temperatures above = 75 °C, where the heat flux through the waste package is higher, the environments are more concentrated, and the material susceptibility to corrosion degradation is highest. The waste package is always hotter than its surrounding environment, probably by several degrees C at > 70 °C.
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INTRODUCTION ABSTRACT The thermal spray coating technique was employed to produce insulated protective coatings (IPC) on 304 stainless steel (SS) surfaces. The electrochemical corrosion potential (ECP) response and flow-assisted corrosion (FAC) rate were then evaluated in high temperature water under various water chemistry conditions. The ECP results clearly demonstrate that the IPC layer created with a powder of yttria-stabilized zirconia (YSZ) restricted oxidant transport to the metal surface, and the ECP remained at <-500 mVsh e in 288°C water containing 700 - 900 ppb oxygen (O2) and no hydrogen (H2). In addition, no significant weight loss of the YSZ coating was measured after a 10 week immersion and the presence of noble metals on the YSZ coating layer did not alter the ECP behavior of the YSZ coated 304 SS. IGSCC of sensitized SS components in boiling water reactors (BWRs) is known to be a major concern [1, 2]. The IGSCC susceptibility of structural materials in BWRs is influenced by the ECP (see Figure 1), which is controlled by the concentrations of oxidizing and reducing species and the hydrodynamic water flow conditions [3, 4]. The ECP can also be affected by the electronic/ionic conductivity of oxide films formed on metal surfaces in aqueous environment [5]. By lowering the ECP of SS below a critical potential (-230 mV) vs. the standard hydrogen electrode (SHE), the susceptibility to IGSCC is markedly reduced. Recently, it has been reported that the 304 SS oxide film consisted of three structures with different particle sizes: the outer oxide layer with a large particle (Ni-enriched, FesO4-type structure), the outer oxide layer with an intermediate-size particle ((z-Fe203), and the inner oxide layer with a very fine grain (Cr-enriched, Fe304 type structure) [6]. Figure 2 shows the surface and cross-section views of oxide film formed on Typ2 304 SS for 3 weeks in 288°C water containing 250 ppb O2. One approach to lowering the ECP is to decrease the dissolved oxidant levels by adding hydrogen in the feedwater, a process known as hydrogen water chemistry (HWC). Large amounts of hydrogen are normally required to sufficiently lower the dissolved oxidant concentration, so that a ECP below -230 mV is attained. However, several side effects of H she 2 16 60 addition have been reported, such as increased N carry-over to the turbine, higher Co deposition rate, high H a cost, etc. In addition, the IGSCC protection potential (-230 mVshe) is difficult to achieve in highly oxidizing and/or high fluid flow regions.
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Stress Corrosion Crack Growth Rate Behavior of Various Grades of Cold Worked Stainless Steel in High Temperature Water
Andresen, Peter L. (General Electric Corp. R&D) | Catlin, William R. (General Electric Corp. R&D) | Young, Lisa M. (General Electric Corp. R&D) | Horn, Ronald M. (GE Nuclear Energy)
ABSTRACT The influence of yield strength on stress corrosion cracking in stainless steel is important because of surface cold work, bulk cold work, weld shrinkage strain, and irradiation hardening. To this end, stress corrosion crack growth rate measurements were performed in high temperature, ultra high purity water on unsensitized stainless steels (and alloy 600) of various grades and compositions as a function of heat, yield strength (cold work), martensite content, corrosion potential, temperature, and hydrogen fugacity. SCC growth rate responded to changes in yield strength, corrosion potential, and temperature - and was substantially independent of the martensite content per se, and the hydrogen fugacity (or hydrogen permeation rate). Unlike commercial heats, a model "stainless steel" alloy containing 5% Si showed high growth rates, little effect of corrosion potential, and little decrease in growth rate with decreasing stress intensity factor. Key Words" stainless steel, stress corrosion cracking, high temperature water, crack growth rate, corrosion potential, water purity, stress intensity factor, weld residual strain, cold work, sensitization. INTRODUCTION Stress corrosion cracking (SCC) has occurred in many materials used in boiling water reactors (BWRs) and pressurized water reactors (PWRs). Despite many observations and common characteristics, SCC is often compartmentalized into small, unique modes with individualized mechanisms and dependencies [1-7]. It is now acknowledged [8] that the crack tip is deaerated and at low potential in all cases (with the increasing adoption ofNobleChem TM in BWRs, even the surface corrosion potential is low), the environmental conditions under which crack advance occurs in BWR and PWR primary systems are very similar [ 1,2,4-7]. Thus, BWRs and PWRs differ primarily in: coolant additives that shift the pH at
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- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (0.66)