This paper summarizes the results from a 4-year program of evaluating the effect of very low levels of chloride (< 5ppb) on the stress corrosion cracking (SCC) behavior of low-alloy pressure vessel steel (A533B) in high temperature water. Three different heats of low alloy steel from reactor vessels were evaluated under both periodical partial unloading condition and constant load/K condition. Effects of chloride (1-5 ppb), stress intensity factor (K), specimen orientation and hydrogen chemistry were studied in detail. It was found that the compositional percentage of sulfur is a weak measure of SCC susceptibility under ppb level of chloride. The result from this program suggests that the chloride level as low as 2-4 ppb can increase the SCC susceptibility of low alloy steel in BWR environment.
The reactor pressure vessel (RPV) is the most critical pressure boundary component in boiling water reactors (BWRs) in terms of plant safety and reliability. Stress corrosion cracking (SCC) of pressure vessel low alloy steel (LAS) in high temperature water has been widely studied since 1970s 1-5. The major factors and SCC susceptibility conditions have been identified and adequately understood with acceptable reproducibility and different mechanistic models. The dissolution of MnS inclusions, intersected and exposed by a growing crack, is a primary factor influencing corrosion fatigue and SCC of carbon steel and low alloy steel. This process is generally exacerbated by the presence of dissolved oxygen which concentrates anions such as HS– and Cl– in the crack tip 1. Higher sulfur steel and higher corrosion potential tend to promote SCC growth in high temperature water.
In addition to the sulfur effect, it has also been well known that there is an important synergy between the crack growth rates, which affects the rate at which the crack intersects new MnS inclusions, and the crack chemistry. This creates a hysteresis in response as loading is decreased vs. increased 6. The primary controlling parameter is the growth rate which determines the MnS intersection rate and the high S crack chemistry. The sulfur content, the crack orientation relative to MnS stringers, and the dissolved oxygen concentration, together influence the crack chemistry, and therefore the crack growth rate of low alloy and/carbon steels.
Materials degradation in light water reactors, particularly of primary pressure boundary components, has had a significant effect on capacity factor. Early degradation issues developed within months of operation – such as bulk cold worked stainless steel in boiling water reactors (BWRs) – and were resolved. Subsequent problems tended to develop on a roughly geometric time scale (e.g., 1.5X), so that new generations of issues surfaced after 1.1, 1.7, 2.5, 3.8, 5.7, 8.5… years. In general, each problem resolution created a sense of optimism that all degradation problems were resolved, when in fact most problems were mitigated but not eliminated, and new problems emerged. In 2002, efforts to anticipate degradation were initiated at the Nuclear Regulatory Commission (NRC)(1) and at the Electric Power Research Institute (EPRI)(2) in the form of proactive materials degradation management, which relied on experience and expert judgment. The systematization of the vulnerabilities unfortunately led to steadily diminished research to address the many gaps. Additionally, the roughly geometric timescale between the appearance of new issues has led some to believe that if we haven’t seen it yet, it probably won’t happen – the same thinking that existed since the inception of light water reactor (LWR) operation. This paper highlights a variety of reasons why historical experience does not provide an adequate basis for anticipating future degradation, and encourages sustained expertise, laboratory capability and plant inspection so that future degradation can be identified and managed. This is especially important as extended lifetimes of 60 – 100 years are considered.
As light water reactor operation continues, there is concern for the effects of thermal aging of structural materials and welds. Cast stainless steels are a particular concern, but stainless steel weld metals are also ‘cast’ and always have ferrite, but may be overlooked. Ferrite is an essential feature of stainless steel welds, where a ferrite level of 5 – 10% is generally targeted but 15% ferrite is sometimes observed. Key concerns include the effect of aging on fracture toughness and on stress corrosion cracking. This paper focuses on the effect of aging on stress corrosion cracking in BWR water.
Reliable operation of existing and new light water reactors over extended lifetimes of 60 – 100 years may be best achieved with improved materials, and this program is designed to evaluate candidate materials for use both in fuel cladding and structural applications. The rapid corrosion of zircaloy fuel cladding at the Fukushima Daiichi plants produced large quantities of hydrogen, which was responsible for explosions that damaged the plants and further hampered the efforts to re-establish adequate cooling to the reactors. Structural materials in current use, including standard austenitic stainless steels for the core shroud and piping in boiling water reactors, have experienced on-going issues with stress corrosion cracking, causing increased inspections, repairs and replacements. Some of the same alloys that are strong candidates for fuel cladding are also good candidates in structural applications. The reduced nickel content of ferritic “stainless steels” translates to lower cost and lower activation, and these alloys also possess superior resistance to radiation damage and stress corrosion cracking. This paper will summarize the data obtained to quantify the SCC growth rate response of a range of ferritic steels being considered as candidate alloys for fuel cladding and structural components.
Light water reactors structural materials such as austenitic stainless steels and nickel based alloys containing 15-18 wt% chromium (Cr) may be susceptible to stress corrosion cracking (SCC), especially in the cold worked condition. Nickel based Alloy 690 or N06690 with 30 wt% Cr, is more resistant to SCC in high temperature water. The purpose of this study was to evaluate SCC susceptibility of commercially available alloys with varying contents of Cr between 18 wt% and 33 wt% and with different microstructural conditions in simulated boiling water reactor (BWR) high temperature water chemistries. Results show that alloys with a uniform equiaxed microstructure and Cr concentrations of 25 wt% and higher showed a larger than 10X reduction in SCC growth rates in the water chemistries tested compared to those with Cr concentrations lower than 25 wt%. Type 310 stainless steel or S31008 with ~25% Cr and with a non-uniform recrystallized grain structure had up to a 10X increase in SCC growth rates compared with the same chemical composition alloy with a uniform equiaxed recrystallized grain structure. Austenitic materials with 25 wt% Cr and higher (e.g., modified Alloy 800 or N08800 and Alloy 33 or R20033 ), appear to be improved alternatives to current BWR structural materials.
Extensive data have been obtained in the last decade show that Alloy 690 and its weld metals inherently have at least some susceptibility to stress corrosion cracking (SCC). Low growth rates are observed under good circumstances, but there are vulnerabilities that can increase the growth rates by 1000X. The primary vulnerabilities evaluated relate to microstructural homogeneity and cold work, and above ~ 20% cold work high crack growth rates are often observed. This sensitivity to cold work is applicable to most, if not all, heats and forms of Alloy 690, although reduced grain boundary carbide coverage reduces the vulnerability. This paper presents on-going work on the effect of cold and hot work, solution annealing and the orientation of the crack plane in relation to cold work on many heats of Alloy 690 ranging from plates to control rod drive mechanism (CRDM) housings. This paper also briefly summarizes extensive testing on Alloy 152 and 52 weld metals, which consistently show very low growth rates in the as-welded condition, or in welds with repairs or with re-fuse layers.
Rapid fracture (or sudden fracture) of various wrought and weld metals in the environment have been observed during compact tension (CT) test in different laboratories. Hydrogen distribution throughout the metal and plastic instability during testing has been proposed as two possible mechanisms that could explain such a sudden or very rapid crack advance. However, no data was available to verify these mechanisms. This paper summarizes the findings of the recent Electric Power Research Institute (1) program that was specifically designed to understand this phenomenon in cold-worked stainless steel and alloy 82 (UNS N06082)/182 (UNS W86182) weld metal. An accelerated testing method for rapid fracture was designed to simulate the inadvertent rapid fracture events observed in the lab. 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. Various formulations for limit load analysis were identified, and applied to determine if plastic instability was consistent with those experimental observations. The results in this program suggest that plastic instability plays a very important role on the occurrence of rapid fracture. The role of dissolved hydrogen, at least under the condition that has been tested in this program, is not a big contributor to the phenomenon.
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 (IASCC) mechanism and for establishing strategies for IASCC mitigation. In this study, experimental crack growth rate (CGR) data of irradiated alloys reported by the university and the Cooperative IASCC Research (CIR) 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 (RCT) specimen under constant K control at 15 MPa·m½was 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.