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ABSTRACT Effects of flow rate on intergranular crack growth in sensitized Type 304 stainless steel (UNS S30400) in distilled water containing 15 ppm or 25 ppm (2.59 × 104 or 4.31 × 104 m) sodium chloride (NaCl) at 250°C were examined using compact tension (CT) specimens under constant loading conditions. On increasing the flow rate, the crack growth rate (CGR) drastically increased, but later decreased to a level that was lower than the initial value. The initial increase in CGR was attributed to an enhanced rate of mass transfer of oxygen to the external surface, where it consumed the current emanating from the crack mouth. However, the subsequent decrease in CGR was attributed to crack flushing, which is a delayed process because of the time required to destroy the aggressive conditions that exist within the crack. Once flushing destroyed the aggressive crack environment, CGR decreased with increasing flow rate. The time over which CGR increased after an increase in the flow rate depended on how fast crack flushing occurred by fluid flow; the higher the flow rate and the greater the crack opening, the faster the crack flushing and the shorter the transition time. Intergranular stress corrosion cracking (IGSCC) in weld-sensitized, Type 304 (UNS S30400)(1) stainless steel (SS) remains a major threat to the integrity of heat transport circuits (HTC) in boiling water reactors (BWR), in spite of extensive research over the last 30 years. Most of the previous IGSCC studies have focused on examining influences of various environmental, mechanical, and metallurgical parameters on fracture processes in austenitic SS in static, high-temperature aqueous environments. However, most structural components in the HTC of BWR are exposed to fast-flowing, high-temperature water. Thus, experimental conditions used in the laboratory often differ markedly from those that exist in operating reactors, and these differences may have a significant impact on the ability to predict the development of damage. Few experimental studies have been reported on effects of fluid velocity on stress corrosion cracking (SCC) of sensitized Type 304 SS in high-temperature
- Research Report > Experimental Study (0.48)
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ABSTRACT INTRODUCTION Cracking incidents have been reported for components that operate in high-temperature water environments such as pressure vessels, steam generators, piping, deaerators, and turbine discs in power-generating plants, the chemical industry, and the paper and pulp industry.1-4 Cracking predominantly has been transgranular and has involved plain carbon steels and low-alloy steels such as SA333Gr6 (UNS K03008)(1), SA106 (UNS K03006), A533B (UNS K13502), A508 (UNS K12539), and deaerator steels. In many cracking incidents and experiments, cracks were found to be associated with inclusions and/or pits, which were assumed to be generated by inclusion dissolution. Szklarska-Smialowska, et al., performed slow strain rate tests (SSRT) on ASTM A302B steel (UNS K12022) in water with a dissolved oxygen concentration from 0 ppb to 10 ppm and at temperatures from 100°C to 260°C.5 At lower temperatures (100°C and 150°C), cracks initiated mainly from inclusions and pits, while at higher temperatures (200°C and 260°C), cracks initiated from apparently smooth surfaces as well as at pits and inclusions. When tested in air at room temperature, specimens cut with their long axis in the short transverse direction of the plate generated cracks caused by mechanical tearing at the nonmetallic sulfide inclusion stringers. Hurst, et al., performed SSRT in water with controlled oxygen content and potential-controlled tests on a reactor pressure vessel steel.6 They found many of the cracks initiated from shallow pits that formed at manganese sulfide (MnS) inclusions outcropping at the surface of the specimen. Inclusions were considered to act as sites for crevices that generated
- Well Completion > Well Integrity > Subsurface corrosion (tubing, casing, completion equipment, conductor) (0.69)
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ABSTRACT Austenitic stainless steels such as type 304 SS (UNS(1) S30400) are subject to a form of intergranular stress corrosion cracking (IGSCC) known as irradiation-assisted stress corrosion cracking (IASCC). Documentation concerning this phenomenon, which occurs in in-core components of type 304 SS and other austenitic alloys present in light water reactors (LWRs), has increased remarkably during the last decade. A recent review article can be found in Reference 1. The ability to reproduce an IASCC failure in the laboratory was first demonstrated in constant extension rate tensile (CERT) tests conducted on irradiated type 304 SS.2 In these tests, the same intergranular mode of cracking and low-ductility characteristic of IASCC were obtained. Also, these tests confirmed that a threshold fast neutron fluence of ~5 x 1020 n/cm2 (E > 1 MeV) was required for IASCC. Specimens with fluences of 1.0 x 1018 and 6.0 x 1019 n/cm2, respectively, showed completely ductile failures, while specimens with fluences in the range from 5.0 x 1020 to 2.77 x 1021 n/cm2 showed varying amounts of IGSCC from 5 to 100%. The present study has used another type of test, a constant load (CL) test, on irradiated type 304 SS. Unlike a CERT test, which maintains a constant extension rate ~0.0254 mm/h (~0.001 in./h) throughout the test, the CL test starts out with zero extension rate (neglecting creep). This rate does not increase until crack initiation and some growth have occurred; then, it increases at an increasing rate until final rupture. Such a difference in extension rates between a CERT and a CL test may affect the rates of passive film breakdown and reformation at the crack tip and, thus, the details of the IASCC mechanism. However, data from both types of tests are useful since it is always desirable to test under conditions similar to those expected in a particular type of service. For example, a CERT test will simulate the 145
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- Well Completion > Well Integrity > Subsurface corrosion (tubing, casing, completion equipment, conductor) (0.64)
- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (0.64)
ABSTRACT Tests were conducted to determine the passive corrosion rate, and the susceptibility of Alloy 22 to localized corrosion and stress corrosion cracking. The passive corrosion rate was found to be on the range of 10 -9 tO 10 -7 A]cm 2. The passive corrosion rate was not strongly dependent on either the solution pH in the range of 2.7 to 8.0 or chloride concentration from 0.028 to 4.0 M. Increasing the temperature from 25 to 95 °C resulted in an increase in the passive current density from 2 × 10 -9 tO 4 X 10 -s A/cm 2. Results from repassivation potential measurements indicate that Alloy 22 was resistant to localized corrosion especially in solutions where the chloride concentration was less than 0.5 M. Similar results were obtained in stress corrosion cracking tests using fatigue precracked wedge loaded double cantilever beam specimens with an initial stress intensity of 34.8 MPa.m 1~2. No stress corrosion cracking was observed after 8 months in 5 percent NaCI at 90 °C. Minor grain boundary attack and limited secondary cracking were observed after an 8 month exposure to 40 percent MgClz at 1 i0 °C. INTRODUCTION The proposed high-level nuclear waste (HLW) repository at Yucca Mountain (YM) Nevada is intended to provide near-complete containment of radionuclides within the waste packages (WPs) for thousands of years and an acceptably low annual dose to the public living near the site. The WP is the primary engineered barrier to the release of radionuclides to the biosphere and the performance of WPs for the initial several thousand years after radioactive waste emplacement is extremely important to protecting public health and safety. Two attributes, slow corrosion of WP materials and engineered enhancements designed to extend WP lifetimes, are identified in the Total System Performance Assessment (TSPA) for the viability assessment (VA) of the YM site ~ as vital to the overall repository performance. As a result, assessment of WP degradation modes and the determination corrosion rates are necessary to evaluate the overall repository performance. The design of the WP has undergone several iterations and many of the WP designs have included corrosion resistant Ni-Cr-Mo alloys. Recently, five enhanced design alternatives (EDA) were considered as a means to prolong WP lifetimes beyond those estimated in the VA 2. In the EDA II design, which is the design of choice, Alloy 22 is selected as the outer container material for the WPs that will be fabricated by shrink fitting the outer container to a 5-cm thick inner container fabricated of types 316 nuclear grade (NG) or 316L stainless steel (SS). The WPs will be enclosed by a self-supported, 2-cm thick Ti-grade 7 mailbox-shaped drip shield with overlapping sections that will extend over the length of the horizontal emplacement drifts. It is assumed that the slow uniform passive corrosion of the 2-cm thick Alloy 22 container will lead to WP lifetimes well beyond 10,000 yr. The inner container is designed to provide sufficient structural strength during the lifetime of the WP to avoid mechanical failure as a result of rock fall, but no performance allocation in terms of corrosion resistance is assigned to this container. It is assumed that Alloy 22 will be immune to localized (crevice) corrosion because a RH greater than 80 percent will be attained only at WP surface temperatures lower than 80 °C due to a combination of low thermal loading and high rates of ventilation during preclosure 3. The objective of this paper is to present the results of experimental investigations on passive dissolution, crevice corrosion, and SCC of Alloy 22 in chloride solutions under a range of environmental conditions including temperature, chloride concentration
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ABSTRACT INTRODUCTION Perhaps the best documented and most fully investigated aspect of stress corrosion cracking (SCC) of sensitized type 304 (UNS S30400)(1) stainless steel (SS) in high-temperature water is that related to the boiling water reactor (BWR) pipe cracking experienced in the United States from about 1980 to 1990. The engineering aspects of remedying this problem have been described well by Jones.1 BWRs use a working fluid of very pure water containing ~ 200 ppb dissolved oxygen, and this environment can lead to cracking in the sensitized regions near pipe girth welds. In investigating the cause of the cracking, it soon was realized that the electrode potential of the steel was raised by the presence of dissolved oxygen and that cracking occurred if the oxygen content was high enough to raise the potential to values more noble than ~ 230 mVSHE.1 Subsequently, it was shown that the presence of oxygen in the water could be simulated by the experimentally easier expedient of applied potential control and that cracks could be arrested or made to grow by appropriate control of the applied potential.2-4 The extensive research on SCC of sensitized type 304 SS in high-temperature water performed in the 1970s has been reviewed by Smialowska and Cragnolino4 and Gordon.5 Because of the importance and complication, this subject has attracted further attention, and many papers have been published since 1980.6-26 Many investigations were associated with determination of the critical (or minimum) potential for IGSCC of sensitized type 304 SS in oxygenated or deaerated water.1-4,7-9,11-12,14-19,21,26 The effects of ionic impurities, especially sulfate (SO42) and chloride (Cl), on SCC of sensitized type 304 SS have been of great concern.3-5,7-8,10-13,15,20-24 This not only is because of the wide application of SO42 as a
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- Facilities Design, Construction and Operation > Pipelines, Flowlines and Risers > Materials and corrosion (0.47)