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Collaborating Authors
Andresen, Peter L.
ABSTRACT Despite the selection of materials that have good corrosion resistance in light water reactor (LWR) environments – austenitic stainless steels, Alloy 600 and their weld metals – there has been a significant incidence of stress corrosion cracking (SCC) in many components in both boiling and pressurized water reactors (BWRs and PWRs). Extending plant lifetime to 60, 80 or 100 years will require superior materials, which can only be identified by knowing their vulnerabilities and probably by improving their fabrication and microstructure. A few categories of superior materials have surfaced such as ~30% Cr, ~9% Fe Alloy 690; ~21% Cr, ~42% Fe Alloy 800; higher Cr and Ni stainless steels; and ~5 – 12% Cr ferritic steels. This paper discusses some of the benefits and vulnerabilities of each category and its variants, presents data showing that intermediate Cr/Fe alloys such as Alloy 800 and Alloy 825 are not sufficiently resistant to SCC, and examines the SCC resistance of this class of material when the Cr level is elevated to ~25% Cr.
- Materials > Metals & Mining > Steel (1.00)
- Energy > Power Industry > Utilities > Nuclear (0.67)
- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (0.87)
- Production and Well Operations > Production Chemistry, Metallurgy and Biology > Corrosion inhibition and management (including H2S and CO2) (0.87)
- Well Completion > Well Integrity > Subsurface corrosion (tubing, casing, completion equipment, conductor) (0.66)
SCC in PWR Water of Alloy 690 and Its Weld Metals
Andresen, Peter L. (GE Global Research) | Morra, Martin M. (GE Global Research) | Ahluwalia, Kawaljit (EPRI)
ABSTRACT In the last decade, data has changed the perception of Alloy 690 and its weld metals from materials that are immune from stress corrosion cracking (SCC) to materials that can exhibit low growth rates under good circumstances, but have vulnerabilities that can increase the growth rates by 1000X. The primary vulnerabilities evaluated relate to microstructural homogeneity and cold work. With sufficiently high cold work, roughly above 20%, dramatically higher crack growth rates are observed. This appears applicable to most, if not all, heats and forms of Alloy 690. This paper summarized on-going work that presents data on the effect of cold work and the orientation of the crack plane in relation to cold work. Various heats of Alloy 690 have been studied, ranging from plates to control rod drive mechanism (CRDM) housings. This paper summarizes extensive testing on Alloy 152 and 52 weld metals, which consistently show very low growth rates apart from one case where 20% additional strain was applied to the weld metal. Other studies in progress have characterized various welds and heat affected zones, and are beginning to evaluate conditions of greater concern, i.e., related to weld repairs and welds with re-fuse layers, both of which increase the residual strain levels.
- Materials > Metals & Mining (0.95)
- Energy > Oil & Gas > Upstream (0.48)
ABSTRACT ABSTRACT This paper summarizes the structural materials used in light water reactors (LWRs), their historical degradation evolution vs. time, improvements and subsequent vulnerabilities discovered, and possible future degradation phenomena over long-term operation. Little was known about high temperature water environments 50+ years ago, and reasonable but inadequate judgments were made on structural materials based partly on simple, short term tests and partly on intuition of low temperature corrosion phenomena. A steady evolution and improvement in structural materials occurred, especially over the first twenty years of LWR experience. Degradation continues and is managed by a combination of inspection, mitigation and replacement, although concerns for surprises remains because of declining expertise, funding and historical conservatism. Materials of higher resistance undoubtedly exist, and some examples are discussed. New forms of degradation and aging synergies that accelerate the kinetics of known degradation phenomena are considered. INTRODUCTION Stress corrosion cracking (SCC) of structural materials has occurred since the early operation of boiling water reactors (BWRs), pressurized water reactors (PWRs) and other light water reactors (LWR) and high temperature water systems [1-6]. In early LWR designs dating back 50+ years, SCC developed within months or years. As the most susceptible materials and conditions were eliminated, the time to cracking was extended incrementally, e.g., to five years, when further problems developed. The original choice of materials came from the U.S. nuclear navy program, which was the source of many of the original engineers who developed commercial nuclear power. Judgment relied on room temperature experience, whereas the material characteristics and, in particular, protective oxides are quite different in water above about 200 °C. There was also insufficient time to develop highly sophisticated tests whose duration is adequate to the needs of a ~40 year lifetime. There are many forms of materials degradation in LWRs, including corrosion/wastage, SCC, fatigue and corrosion fatigue, aging (phase formation, ordering, etc.),
- Well Completion > Well Integrity > Subsurface corrosion (tubing, casing, completion equipment, conductor) (1.00)
- Production and Well Operations > Production Chemistry, Metallurgy and Biology > Corrosion inhibition and management (including H2S and CO2) (1.00)
- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (1.00)
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)
ABSTRACT: This paper describes the facilities for conducting crack growth rate (CGR) tests on neutron-irradiated alloys in LWR environments. A set of procedures were developed for using the facilities to test CGRs of neutron-irradiated stainless steel in the Irradiated Materials Testing Complex (IMTC) at the University of Michigan. The DCPD technique was used to monitor the evolution of crack growth under various loading and environmental conditions throughout a CGR test up to 3127 hours. The high resolution of DCPD, the application of a correction factor to crack length, and dK/da control when changing K improved the accuracy of crack length monitoring. A procedure for mounting a neutron-irradiated RCT sample in the hot cell using remote manipulator operation was also developed. Precision design of fixtures helped to overcome the difficulties of spot-welding thin DCPD wires using manipulators. On-the-fly environmental change from BWR NWC to BWR HWC, and then to PWR water at 288°C and 320°C was successfully conducted, and constant K control was maintained throughout the test. INTRODUCTION Irradiation-assisted stress corrosion cracking (IASCC) has been recognized as a pervasive degradation mode for structure materials in light water reactors (LWRs). IASCC of some small components occurred initially. In the last ~20 years, the significance of IASCC has been reported in structural components such as BWR core shrouds 1, 2 and PWR baffled bolts. 3, 4 Studies on IASCC have focused on the effect of irradiation on water chemistry, mechanical properties of irradiated alloys, and microstructructual and microchemical evolution of irradiated alloys. 5 The susceptibility of IASCC was investigated by conducting constant extension rate test (CERT) on proton irradiated 5, 6 or neutron irradiated stainless steels 5, 7-9 in simulated BWR and PWR water. Proton irradiated samples worked as well as surrogates for neutron-irradiated samples for studying the crack initiation behavior.
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ABSTRACT: The effect of microstructure, stress intensity factor, corrosion potential and water purity on stress corrosion crack growth rate behavior of Alloy X-750 was investigated in 288 ?C water. This material was provided by a utility in the form of an unused stabilizer support bracket. Alloy X-750 has exhibited SCC in field components, and this study was designed to examine its microstructure and SCC response in some detail to determine the suitability of Alloy X-750 for long term reliable service in BWRs. INTRODUCTION Most structural materials in boiling water reactors (BWRs) and pressurized water reactors (PWRs) have exhibited stress corrosion cracking (SCC). While the crack growth behavior of various grades of stainless steel, alloy 600, alloy 182 weld metal, and low alloy and carbon steels have been extensively investigated [1-7], there is only limited stress corrosion crack growth rate data on Alloy X-750 in high temperature, pure water [8-10], especially in processing forms and heat treatments relevant to plant components. While focused on BWR environments, these studies are strongly relevant to PWR environments, as the crack tip is deaerated and at low potential in all cases (with the increasing adoption of NobleChem? in BWRs, even the surface corrosion potential is low). Thus, the crack tip conditions in BWR and PWRprimary systems are closely related [2-4,11-13]. BWRs and PWRs differ primarily in temperature (274 ?C, to up to 338 ?C in PWRs), H2 fugacity (10 – 150 ppb vs. ? 3000 ppb in PWRs) and coolant additives that shift the pH at temperature from 5.6 to 6.8 – 7.4 in PWRs. The best estimates account for the Ni/NiO phase transition by maintaining a fixed potential difference relative to Ni/NiO line and show an activation energy of 134 kJ/mole (32 kcal/mole) for Ni alloys in deaerated water [14,15].
- Research Report > Experimental Study (0.66)
- Research Report > New Finding (0.46)
SCC Mitigation By Electrocatalysis
Andresen, Peter L. (GE Global Research Center ) | Kim, Young J. (GE Global Research Center )
ABSTRACT: SCC growth rates are strongly influenced by water chemistry parameters, especially when crack chemistry can be concentrated from differential aeration or thermal gradients or boiling. Mitigation of the effects of the high corrosion potential associated with oxidants is most efficiently achieved by electrocatalysis, which requires that there be a stoichiometric excess of reductants over oxidants. Mechanisms and criteria for effective SCC mitigation are summarized, with particular focus on the critical location for the catalyst in a crack and experimental support for these concepts. Optimization of electrocatalysis by a continuous electrocatalytic process is described, for example where Pt is injected at levels of 0.002 to 0.05 ppb. INTRODUCTION Stress corrosion cracking (SCC) of structural materials has occurred in boiling water reactors (BWRs) and other high temperature water environments. The crack growth behavior has been extensively evaluated for various stainless steels, Alloy 600, Alloy 182 / 82 weld metal, and low alloy and carbon steels, under unirradiated and irradiated conditions [1-10]. From early studies it was clear that corrosion potential has a very strong influence on SCC in high temperature water (Figure 1), and various SCC mitigation methods to decrease the corrosion potential were developed. This paper summarizes the electrochemical kinetics mechanism by which Pt group metals act in high temperature water. The kinetic requirement in typical gas phase catalysis, where reaction rates are very high, are more stringent than 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, H2O2, and H2). Thus, dilute noble metal alloys (e.g., stainless steel containing 0.1% Pd, Pt, etc.) and/or low surface loadings also exhibit catalytic behavior. The injection of some H2 in a BWR also lowers the oxidant concentration, further reducing the kinetic requirements for the catalysts.
- Production and Well Operations > Production Chemistry, Metallurgy and Biology > Corrosion inhibition and management (including H2S and CO2) (0.66)
- Well Completion > Well Integrity > Subsurface corrosion (tubing, casing, completion equipment, conductor) (0.46)
- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (0.46)
SCC of Alloy 690 And Its Weld Metals
Andresen, Peter L. (GE Global Research Center ) | Morra, Martin M. (GE Global Research Center ) | Ahluwalia, Kawaljit (EPRI)
ABSTRACT: Extensive SCC growth rate testing of Alloy 690 base metal, HAZ and weld metal was performed in representative PWR primary water at 290 to 360 ?C. Intergranular cracking was observed in all materials. Growth rates as high as 1.2 x 10–6 mm/s were observed in the S-L orientation with microstructural banded material after cold rolling or forging to align the planes of banding, rolling and cracking. However, not all banded material has exhibited such high growth rates. Growth rates on homogeneous Alloy 690, including extruded CRDM tubing, often showed growth rates in the range of 2 – 8 x 10–8 mm/s in cold worked condition and an S-L orientation. Crack growth rates in some Alloy 690 tests were in the range of 1 to 10 x 10–9 mm/s, primarily in orientations other than S-L. For cracks aligned along the HAZ, growth rates as high as 1.2 x 10–8 mm/s were observed. Alloy 152/52/52i weld metals always exhibited low growth rates, apart from a weld that was further cold worked by 20%, which grew at 7 x 10–9 mm/s. INTRODUCTION Most structural components in light water reactors have experienced stress corrosion cracking (SCC) [1- 10], although the rate of crack growth varies immensely. The resistance to SCC of newer materials, such as Alloy 690 is significantly improved, but SCC immunity appears not to exist [1-2,8-10]. Combinations of materials, environments, and stressing conditions – such as unsensitized stainless steel and Alloy 690 in boiling water reactor (BWR) or pressurized water reactor (PWR) water – that were once generally viewed as immune to SCC have now been shown to be susceptible [1,2,5-7]. This is primarily because historical testing was not adequately sensitive or sophisticated. Figure 1 shows that after carefully transitioning from fatigue to constant load, sustained and
- Materials > Metals & Mining > Steel (0.36)
- Energy > Power Industry > Utilities > Nuclear (0.35)
- Energy > Oil & Gas > Upstream (0.34)
ABSTRACT: Rapid fracture (or sudden fracture) of various wrought and weld metals in the environment has been observed during compact tension (CT) testing in different laboratories. The presence of hydrogen distribution throughout the metal has been proposed as the most possible mechanism to explain such a sudden or very rapid crack advance. This paper summarizes those findings and reports on controlled experiments designed to understand this phenomenon in cold-worked stainless steel and Alloy 82/182 weld metal (UNS W86182/UNS N06082). SEM characterization of previous fractured specimens shows that rapid fracture usually produces a ductile dimple fracture morphology, although such fracture occurred at K values as low as 41.8 MPa_m. The effects of dissolved H2 and/or O2 concentration, stress intensity factor (K), and the change in K on the occurrence of rapid fracture were investigated using CT specimens in high temperature water. The tests were designed to sustain stress corrosion crack (SCC) growth over sufficient time to allow hydrogen to permeate throughout the specimen, then slowly increase K to intentionally reproduce the rapid fracture and evaluate the role of hydrogen on this behavior. The possibility of developing a specific protocol or guideline to evaluate this phenomenon in a fast and repeatable way was also evaluated. INTRODUCTION Rapid fracture of structural material, which can also be called sudden fracture, has been observed as very rapid crack advance on compact tension (CT) specimen during SCC test in high temperature water. A growing number of laboratory observations show this phenomenon on various structural materials at K values lower than observed in air. There are more than two dozen incidents from at least five different laboratories. The materials include stainless steels, nickel alloys, and nickel alloy weld metals. Rapid fracture can take place in both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) water chemistries.
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- Materials > Metals & Mining > Nickel (0.55)
INTRODUCTION ABSTRACT: Austenitic stainless steels such as type 304 and 316 are used as core internals components in light water reactors. Under certain operation conditions the austenitic stainless steels may be susceptible to environmentally assisted cracking, more specifically, irradiation assisted stress corrosion cracking. Irradiation also could produce hardening and alter the local chemistry of the austenitic alloys, which would increase their susceptibility to environmental cracking. Ferritic stainless steels are less susceptible to irradiation damage including void swelling. Ferritic stainless steels also offer desirable higher thermal conductivity and lower thermal expansion coefficient. Little is known however about the stress corrosion cracking behavior of ferritic steels in high temperature water. Crack propagation rate studies were conducted using four types of ferritic steels in high purity water at 288°C containing dissolved oxygen or dissolved hydrogen. Results show that ferritic steels (5% Cr, 9% Cr, 12%Cr and 17% Cr) are notably more resistant to environmental cracking than the austenitic materials. Austenitic stainless steels such as types 304 and 316 are highly susceptible to stress corrosion cracking in chloride containing environments. 1 Ferritic stainless steels such as types 405 and 430 are highly resistant to SCC in hot chloride solutions.1,2 Austenitic stainless steels (SSs) core internals components are susceptible to irradiation assisted stress corrosion cracking (IASCC) during service in nuclear power plants light water reactors. 3,4,5 One of the effects of irradiation is the hardening of the SS due to modifications in the dislocation distribution in the alloy. 6,7,8 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 (SCC).
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