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A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components

In the United States currently there are approximately 104 operating light water reactors. Of these, 69 are pressurized water reactors (PWRs) and 35 are boiling water reactors (BWRs). In 2007, the 104 light-water reactors (LWRs) in the United States generated approximately 100 GWe, equivalent to 20% of total US electricity production. Most of the US reactors were built before 1970 and the initial design lives of most of the reactors are 40 years. It is expected that by 2030, even those reactors that have received 20-year life extension license from the US Nuclear Regulatory Commission (NRC) will begin to reach the end of their licensed periods of operation. For economic reason it may be beneficial to extend the licenses beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean, and economic electricity without significant green house gas emissions [1]. However, environmental assisted damage and aging issues are some of the major concerns for long-term viability of these nuclear reactors. Despite regular maintenance and tightly regulated operating procedures, aging related failures do occur in US nuclear power plants (NPP).

Different forms of aging might be active in the NPP components [2-7]. They include pure irradiation-induced hardening and softening, irradiation-induced swelling, phase transformation, creep, thermal aging such as thermal hardening and softening of material properties, thermally induced high-cycle and low-cycle fatigue, and high-cycle mechanical fatigue due to flow- induced vibration, chemical corrosion related damage such as flow-assisted general corrosion, crevice corrosion, and stress corrosion cracking (SCC). These mechanisms can act individually or act in combination to accelerate the aging processes. For example, flow accelerated corrosion (FAC), corrosion fatigue (CF), and irradiation-assisted stress corrosion cracking (IASCC) can act in combination with each other to magnify their individual effect. Different NPP components are subjected to different damage mechanisms depending on the type of material used, the way it is manufactured, and the exposure environments. For example, for long-term operation the reactor pressure vessel (RPV) may be subjected to irradiation-induced hardening, swelling, phase transformations, and creep. In addition, reactor pressure vessels may be subjected to fatigue damage. Although fatigue is a degradation mechanism that is second in importance to radiation embrittlement for PWR RPV, it may also affect BWR RPV structural integrity. For example, BWR vessels may be subjected to both high-and low-cycle fatigue [2]. US BWR vessels are designed for low-cycle fatigue based on classical S~N curves, which are based on data generated from in-air fatigue tests. However, research results show that there is an effect of high- temperature (200 to 300°C) oxygenated water (typical BWR environment) on fatigue strength of low alloy steel [8]. This questions the long-term performance of US BWR vessels, which are mostly built using low alloy steel. In addition to the main shells, both the PWR and BWR RPVs contain various nozzles and penetrations. Many of these nozzles are joined to the RPV by dissimilar metal welds that may be affected by stress corrosion cracking and/or corrosion fatigue [2].

Within the pressure vessels, there are many internal structures that support the reactor core,maintain fuel assembly alignment, etc. These internal structures are not only subjected to reactor coolant water chemistry, but are also exposed to higher temperature and higher irradiation dose compared to the RPV shell. Major possible degradation mechanisms for these internal components are irradiation embrittlement and associated IASCC. In addition, high-cycle fatigue due to flow-induced vibration and low-cycle thermal fatigue are potential causes of aging for long term operation. Thermal embrittlement may also affect some of the internal cast stainless steel components. For example, although most of the PWR internal components are made from stainless steel, some, such as Combustion Engineering design upper guide structure assembly shrouds are made from cast steel grade CF-8. These parts may be subjected to thermal embrittlement and need to be investigated for long term NPP operation.

In addition to the RPV and internal structures, the reactor cooling system (RCS) pipes are part of the reactor coolant pressure boundary whose structural integrity affects the overall functionality of NPP. RCS cooling system pipes include hot leg and cold leg pipes and steam generator tubes for PWR plants and steam, feed water and recirculation pipes for BWR plants, all of which are critical for the overall safety of the reactors. Stress corrosion cracking is a major issue for RCS system pipes particularly in the weld regions where it is connected to RPV nozzles through safe ends. Also, SCC is a major issue for steam generator tube integrity in many US PWRs. The primary cause of SCC is the residual stress created in the component during manufacturing or fabrication processes. For example, high residual stresses are generated during welding of dissimilar metal joints that connect the pipes to RPV shell. Similarly, high residual stresses are generated in steam generator tubes U-bends during its forming process. In addition to SCC, pressure and thermal stress-induced low-cycle fatigue is also a major concern for the RCS. Pressure and thermal stress are created during system transients including heat up and cool down that could cause low-cycle fatigue damage. This fatigue along with SCC leads to corrosion fatigue. For example, SCC/CF can occur in PWR coolant systems nozzles, dissimilar metal welds, and elbows [2]. This review report presents information related to SCC/CF in reactor coolant system piping and weld.