Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments (original) (raw)
Related papers
Crack growth rates of nickel alloy welds in a PWR environment
2006
In light water reactors (LWRs), vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking. A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. A program is being conducted at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated LWR coolant environments. This report presents crack growth rate (CGR) results for Alloy 182 shielded-metal-arc weld metal in a simulated pressurized water reactor (PWR) environment at 320°C. Crack growth tests were conducted on 1-T compact tension specimens with different weld orientations from both double-J and deep-groove welds. The results indicate little or no environmental enhancement of fatigue CGRs of Alloy 182 weld metal in the PWR environment. The CGRs of Alloy 182 in the PWR environment are a factor of ≈5 higher than those of Alloy 600 in air under the same loading conditions. The stress corrosion cracking for the Alloy 182 weld is close to the average behavior of Alloy 600 in the PWR environment. The weld orientation was found to have a profound effect on the magnitude of crack growth: cracking was found to propagate faster along the dendrites than across them. The existing CGR data for Ni-alloy weld metals have been compiled and evaluated to establish the effects of key material, loading, and environmental parameters on CGRs in PWR environments. The results from the present study are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of the specific Ni-alloy weld to environmentally enhanced cracking. iv Intentionally Left Blank Foreword v This report presents crack growth rate data and the results of the corresponding fracture surface and metallographic examinations from cyclic loading and primary water stress-corrosion cracking (PWSCC) tests of two nickel-base Alloy 182 (A182) weldments, which are typical of those used in vessel penetrations and piping butt welds in nuclear power plants. The effect of crack orientation with respect to dendrite orientation is the most significant variable investigated in this study. However, this report also includes a review of data from several laboratories, which describes the effects of material composition, loading characteristics, and chemistry of the aqueous environment. The main conclusion is that the PWSCC growth rates described for A182 specimens in this report are comparable to the crack growth rates that characterize the performance of Alloy 600 (A600). This report is the first in a series documenting the results of crack growth rate testing in vessel head penetration materials, focusing on the weld metals, A182 and A152, and including results of some tests of the base metals, A600 and (eventually) A690. The results presented in this report were obtained in tests of a laboratory-fabricated, shielded metal arc welding deposit of A182. Testing of A182 weldments continues at Argonne National Laboratory, and substantially more crack growth rate results are anticipated in the next two years. The impetus for this research on PWSCC comes from User Need Request NRR-2002-018, submitted by the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation. This topic may be an especially important consideration in the review of license applications, as well as the disposition of relief requests pertaining to flaw evaluations for vessel penetration and piping butt welds. The data on cyclic loading effects are commonly used in the fatigue analyses that are required for flaw evaluations completed in accordance with the requirements set forth in Section XI, IWB-3660 and Appendix O, of the Boiler and Pressure Vessel Code promulgated by the American Society of Mechanical Engineers.
Crack Growth Rates of A182 and A82 Alloys From the V.C. Summer Reactor Vessel Nozzle-to-Pipe Weld
Volume 1: Codes and Standards, 2005
The Ni-base alloys used as construction material in light water reactors (LWRs) have experienced stress corrosion cracking (SCC). Although SCC of wrought Ni-base Alloy 600 has been observed in operating plants for many years, until recently, the weld metal Alloys 82 and 182 used with Alloy 600 environmentally assisted cracking has not been widely observed in the field. However, laboratory tests indicate that in PWR coolant environments, the SCC susceptibility of Alloy 182 may be greater than that of Alloy 600, and that of Alloy 82 may be comparable to Alloy 600. This paper presents crack growth rate (CGR) results for Alloys 182 and 82 from the reactor vessel nozzle-to-pipe from the V. C. Summer plant under both constant and cyclic load. The tests were conducted on ½-T compact tension specimens in a simulated PWR environment at 320°C. Crack extensions were measured by DC potential drop measurements. Characterization of the material microstructure and is described. The SCC growth rates are compared with the existing CGR data for Ni-alloy welds to evaluate the effects of alloy type, weld microstructure, and stress intensity factor K on CGRs. The cyclic CGRs for these alloys are compared with CGRs that are expected for Alloy 600 in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions for significant environmental enhancement in growth rates. A detailed characterization of the fracture morphology is also presented.
Materials Characterization, 2016
Ni-based weld alloys 52, 52M and 152 are extensively used in repair and mitigation of primary water stress corrosion cracking (SCC) in nuclear power plants. In the present study, a series of microstructure and microchemistry at the SCC tips of these alloys were examined with scanning electron microscopy (SEM), electron backscatter diffraction (EBSD), transmission electron microscopy (TEM), energy-dispersive X-ray spectroscopy (EDS), scanning transmission electron microscopy (STEM) and energy filtered transmission electron microscopy (EFTEM). The specimens have similar chemical compositions and testing conditions. Intergranular (IG) and transgranular (TG) SCC was observed in all of them. The cracks were filled with nickel-oxides and partial precipitations of chrome carbides (CrCs), niobium carbides (NbCs), titanium nitrides (TiNs) and silicon carbides (SiCs), while iron (Fe) was largely dissolved into the solution. However, the crack densities, lengths and distributions were different for all three specimens.
CORROSION, 2008
Thermal aging and consequent embrittlement of materials are ongoing issues in cast and duplex stainless steels. Spinodal decomposition is largely responsible for the well-known "475°C" embrittlement that results in drastic reductions in ductility and toughness in cast materials. This process is also operative in welds in cast or wrought stainless steels where delta ferrite is present. While the embrittlement can occur after several hundred hours of aging at 475°C, it can also occur at lower temperatures where ductility reductions have been observed after tens of thousands of hours at 300°C. The effect of thermal aging on mechanical properties, including tensile, toughness, fatigue, and static crack growth, has been investigated at room temperature and in 288°C high-purity water simulating boiling water reactor (BWR) operating conditions. The measurements of tensile, microhardness, and Charpyimpact energy show an increase in strength and a decrease in impact energy after aging for up to 10,000 h at 430°C and 400°C. Stress corrosion crack (SCC) growth rates have been measured for as-welded and 5,000-h/400°C aged weld metal at 288°C in high-purity water containing 300 ppb of oxygen. Fracture toughness (J IC) have been measured in the 5,000-h/ 400°C aged weld metal and estimated in the other conditions. Crack growth rates for material in the as-welded and aged metal for 5,000 h at 400°C have been measured and are generally within the scatter band for wrought material, although the aged material data fall at the high end. Unusual in situ unstable fracture behavior has been experienced for material that contains an SCC "precrack" at toughness values significantly below (<50%) the room temperature fracture toughness. In situ fracture toughness with a fatigue precrack is still signifi cantly below the air values. This behavior, termed "environmental fracture," requires further investigation.
Journal of Technology Innovations in Renewable Energy, 2014
One of the main degradation mechanisms which cause risks to safety and reliability of pressurized water nuclear reactors is the primary water stress corrosion cracking (PWSCC) in nickel alloys, such as Alloy 600 (75Ni-15Cr-9Fe), and its weld metal Alloy 182 (67 Ni-15Cr-8Fe). It can appear at several reactor nozzles dissimilarly welded with Alloys 182/82 between steel ASTM A-508 G3 and stainless steel AISI316L, among others. The hydrogen which is dissolved to primary water to prevent radiolysis, can also have influence on the stress corrosion cracking behavior. In this article one departs from a study of Lima based in experimental data obtained from CDTN-Brazilian Nuclear Technology Development Center, in slow strain rate test (SSRT). It was prepared and used for tests a weld in laboratory, similar to dissimilar weld in pressurizer relief nozzles, operating at Brazilian NPP Angra 1. It was simulated for tests, primary water at 325 o C and 12.5 MPa containing levels of dissolved hydrogen: 2, 10, 25, and 50 cm 3 STP H2/kgH2O. The objective of this article is to propose an adequate modeling based on these experimental results, for PWSCC crack growth rate according to the levels of dissolved hydrogen, based on EPRI-MRP-263 NP. Furthermore, it has been estimated the stress intensity factor applied for these tests: according with these, some another models described on EPRI-MRP-115, and an USNRC Technical Report, have been tested. According to this study, CDTN tests are adequate for modeling comparisons within EPRI and USNRC models.
Stress corrosion cracking initiation in alloys 600 and 182
1993
Wrought Alloy 600 and Alloy 182 weld metal were tested for stress corrosion crack initiation in a BWR test loop. For the tests a modified bolt loaded compact tension specimen was developed, in which a U-notch replaced the conventional crack notch. The stress state close to notches of various radii was analyzed. Specimens with various notch radii were loaded to the same displacement at the bolt. A few control specimens with conventional crack notches were included. In Alloy 600 there was very little crack initiation, if any. However, in Alloy 182 weld metal there was crack initiation in a considerable fraction of the tested specimens. Comparison to other data from the same project shows that materials which had high crack propagation rates also showed high susceptibility to crack initiation. The appearance of a few barely initiated cracks indicated that local chemical attack on weld dendrite boundaries was the major factor causing initiation of cracks in Alloy 182. (Less)
Journal of Technology Innovations in Renewable Energy, 2014
Dissimilar welds (DW) are normally used in many components junctions in structural project of PWR (Pressurized Water Reactors) in Nuclear Plants. One had been departed of a DW of a nozzle located at a Reactor Pressure Vessel (RPV) of a PWR reactor, that joins the structural vessel material with an A316 stainless steel safe end. This weld is basically done with Inconel/Alloy 182 with a weld buttering of Inconel/Alloy 82. It had been prepared some axial cylindrical specimens retired from the Alloy 182/A316 weld end to be tested in the slow strain rate test machine located at CDTN laboratory. Based in these stress corrosion susceptibility results, it was done a preliminary semi-empiric modeling application to study the failure initiation time evolution of these specimens. The used model is composed by a deterministic part, and a probabilistic part according to the Weibull distribution. It had been constructed a specific Microsoft Excel worksheet to do the model application of input data. The obtained results had been discussed according with literature and also the model application limits.
Hot cracking and environment- assisted cracking susceptibility of dissimilar metal welds
The operating experience of major nuclear power plant (NPP) pressure boundary components has recently shown that dissimilar metal weld joints can markedly affect the plant availability and safety because of increased incidences of environment-assisted cracking or primary water stress corrosion cracking (EAC, PWSCC) of alloy 600 and corresponding weld metals (alloys 182/82). Alloy 690 and associated weld metals (alloys 152/52) are widely used for repair and replacement of the affected thick-section components. The selection of new materials relies mainly on laboratory results and short-term service experience. The long-term behavior of these materials and their performance in the plant has still to be demonstrated. Weldability and susceptibility to hot cracking of the studied nickel-base materials was evaluated based on the Varestraint test results obtained with weld metals of different chromium contents. The microstructures and microchemistry of the multi-pass nickelbase alloy welds was studied by FEG-SEM/EDS techniques and were found to be very different from those of the wrought and recrystallized nickel-base materials. Additionally, the weld residual stresses were measured and analyzed by a novel Contour method suitable for through-thickness residual stress determination. The studied nickel-base material welds were exposed to doped steam test environment and crack initiation susceptibility of them was studied. The results showed the markedly higher susceptibility to EAC of alloy 182 weld metal as compared to the other studied alloys, i.e., alloy 152, 52 and 82. * Crack mouth opening displacement (CMOD). * −∆(2 * δ) or-2∆δ in Figure 3.3. ** Strain relaxation in outer surface, when the specimens were unloaded after the autoclave exposures.
Nuclear Engineering and Technology, 2015
Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.