Study on Stress Corrosion Cracking Behavior of Stainless Steel for Nuclear Power Plant in High Temperature and High Pressure Water

With the rapid development of industry, the demand for energy is increasing, and traditional coal-fired power plants cause serious environmental pollution. With the international community paying more and more attention to greenhouse gas emissions and climate warming, China has listed the development of nuclear power as one of the important measures to solve environmental problems. The Chernobyl nuclear accident in 1986 and the Fukushima nuclear accident in Japan in 2011 triggered an explosion, resulting in the large-scale leakage of radioactive nuclear materials, and the global nuclear safety has become increasingly important. Šobrīd, the types of nuclear power plants operating in the world are divided into light water reactors and heavy water reactors. Light water reactors include pressurized water reactors (PWRs) and boiling water reactors (BWRs). Corrosion of structural materials, especially stress corrosion cracking (SCC), is a major problem affecting the safety of equipment and pipelines in nuclear power plants. In order to improve the corrosion resistance of nuclear power equipment, most of the structural materials used in water-cooled nuclear reactors are nickel-based alloys and austenitic stainless steels with good corrosion resistance and mechanical properties. The superior corrosion resistance is mainly due to the formation of chromium-rich oxide film (passivation film) on the material surface in the corrosive medium.

The service environment of nuclear metal materials is usually high temperature and high pressure water environment, accompanied by a certain degree of radiation. The harsh service environment and long-term exposure make the structural materials of nuclear power plants in a corrosive state. SCC in high temperature and high pressure water refers to the process of crack initiation, propagation and cracking caused by local defects under the joint acceleration of sensitive structural materials, corrosive media and stress. Once SCC is initiated, it will expand rapidly on equipment and materials, leading to component failure, coolant leakage, and even unit shutdown, which directly threatens the safe operation of nuclear power plants. Tāpēc, the stress corrosion of stainless steel in the high temperature and high pressure water environment of nuclear power plants has become the focus of attention at home and abroad, especially in the past decade, with the improvement of the application and demand for clean energy and the increase of the severity of the requirements for safe operation of nuclear power plants, the international research on stress corrosion of stainless steel materials in nuclear power plants is in a rapid growth period.

China’s nuclear power industry started relatively late. Šobrīd, most of the nuclear power plants in operation and construction are imported reactors and use pressurized water reactors. The relevant technology is immature, the basic research on water chemistry is lacking and weak, and the practical experience is few. The structural materials used in nuclear power plants are mainly 304 and 316 stainless steel, nickel base alloy 600 and 690, welding metal nickel base 52/152 alloy and carbon steel. The research on corrosion prevention and radiation protection of structural materials of nuclear power plants in China has explored and applied the water chemistry technology suitable for domestic nuclear power plants while drawing on the practical experience of foreign countries and combining with the actual situation in China.

1 SCC of nuclear-power stainless steel

Due to its good plasticity, corrosion resistance and processing performance, stainless steel is widely used in the main equipment, pipes and welds of the PWR nuclear island. It is mainly used in the reactor pressure vessel surfacing layer, internals shroud bolts, push rod drive mechanism, main reactor coolant system pipes and other parts. In the 1970s, limited examples of stress corrosion were found on stainless steel materials in the high strain hardening region of the main system of the pressurized water reactor. The cracking caused by pure mechanical stress is different from SCC in normal temperature medium. When stainless steel is used in the high temperature and high pressure water environment of nuclear power plant, it will crack even under the condition of extremely low stress. The number of cracks is small, the depth is deep, the width is narrow, and the direction is basically perpendicular to the stress direction. SCC can be transgranular (TGSCC) or intergranular (IGSCC). Tāpēc, it is of great significance to analyze the influence of different material processing processes and water chemical environmental parameters on the stress corrosion behavior of stainless steel materials in high temperature and high pressure water of nuclear power plants, and the interaction and synergy of materials, environment and stress strain.

2 Stress corrosion test method

2.1 Common SCC sensitivity test methods

SCC in structural materials of nuclear power plants will cause problems such as shutdown and maintenance. Failure to find or handle it properly will directly affect the safe and stable operation of nuclear power plants. Domestic and foreign scholars have studied the stress corrosion cracking behavior, crack initiation and crack growth rate (CGR) of stainless steel materials in different environments by different test methods.

In combination with the standard GB/T15970-2018 and ASTM E399, stress corrosion test specimens include smooth specimens, notched specimens and pre-cracked specimens, and the loading methods include constant displacement, constant load and slow strain rate. The constant displacement method applies a constant total displacement to the metal material through the fixture or bolt before the experiment, often including bending specimen, U-shaped specimen, C-shaped specimen, etc. This method is simple in loading mode, cheap in fixing the fixture, and suitable for the change of specimen size in a wide range, but the corresponding stress cannot be accurately quantified, and the analysis of the corresponding stress state is not clear. The slow strain rate test can simplify the application and calculation of stress, and make the specimen completely fracture to determine some parameters to evaluate the SCC sensitivity of the material. Tomēr, the equipment is relatively complex, and there are many factors affecting the determination of the strain rate value. Compared with the bending specimen, it requires a thicker binding frame and loading method. Modern analytical and testing methods such as electrochemical noise technology can provide in situ, continuous and nondestructive monitoring of local corrosion initiation and development. The stress corrosion performance of metal materials is closely related to the material structure, stress level and corrosion medium. When evaluating the stress corrosion sensitivity, appropriate stress corrosion test methods and sample types should be selected. Different test methods may lead to different test results.

2.2 Test method for SCC crack initiation and propagation rate

The metal stress corrosion test method in the above standards is generally applicable to the test of SCC sensitivity under conventional conditions, and also provides support and reference for the test in the special environment of nuclear power, such as SSRT test, U-bend test, C-ring test, etc; The compact tensile (CT) specimen can be combined with the DC potential drop method (DCPD) to measure the crack growth length in situ to determine the crack growth rate.

In recent years, in view of the particularly harsh high temperature and high pressure water environment of nuclear power plants, China has designed special test methods and developed relevant group standards, such as T/CSTM 00080-2019 on crack initiation test, T/CNS 5-2018 on crack growth rate test, etc. The group standard is determined in combination with the existing standards at home and abroad as well as domestic test methods, technical documents and practical experience, providing support for the smooth, safe and effective conduct of sample loading, online monitoring of water chemical parameters, real-time control and stress corrosion test in high temperature and high pressure water environment.

2.3 Electrochemical test method

In addition to directly testing SCC behavior of metal materials, corrosion electrochemistry is also an important method to evaluate metal corrosion resistance, measure corrosion rate and study corrosion mechanism. China has also formulated relevant standards, such as GB/T 24196-2009, T/CNS 6-2018 and T/CNS 3-2018.

3 Factors affecting SCC in high temperature and high pressure water

The stress corrosion behavior of stainless steel in high temperature and high pressure water environment of nuclear power plant is subject to the comprehensive influence of various factors, mainly including material factors (surface treatment, cold processing, heat treatment process, etc.), mechanical factors (yield strength, residual stress, stress intensity factor, load, etc.) and hydrochemical environment (temperature, pH, anion, dissolved oxygen, etc.).

3.1 Material factors

3.1.1 Surface treatment

Material surface defects and scratches caused by processing are unavoidable during operation. Through slow strain rate test (SSRT), Scenini et al. pointed out that surface treatment in high temperature coolant plays an important role in SCC crack initiation. Compared with mechanical processing, the surface of 304L stainless steel sample polished by oxide suspension (OPS) δ The area near the ferrite/austenite interface is more susceptible to SCC, which makes it more sensitive to stress corrosion. In the simulated water environment of PWR primary circuit, transgranular cracks usually form on the surface of machined samples, which has a great relationship with machining marks. Tomēr, there are only a few transgranular cracks on the surface of the well polished material, and the crack morphology is mainly intergranular.

As a widely used surface strengthening process, shot peening uses shot particles to bombard the material surface and implant residual compressive stress, which can offset part of the tensile stress on the surface of the heat transfer tube, improve the fatigue strength of the workpiece, and significantly reduce the stress corrosion sensitivity of stainless steel. Shot peening can affect the macroscopic state, microstructure, hardness, residual stress, martensitic transformation of the material surface, and then affect the stress corrosion sensitivity of the material. Laser shot peening technology has no rebound medium and reaction force during operation, and there is no residue that affects the normal operation of equipment. It can form a deeper compressive stress layer than ordinary shot peening, and does not damage the surface of components. It has obvious strengthening effect and operability. It is considered as a technology that can be applied in the field of nuclear power and has broad application prospects.

3.1.2 Cold working

Cold working in the process of processing, installation and manufacturing of nuclear power equipment will change the internal microstructure of the material. For example, the bending, welding, grinding, stamping and other processes of austenitic stainless steel will cause plastic deformation of the material, dislocation and point defects will cause lattice sliding, grain boundary orientation, dislocation density and other changes. The change of local mechanical properties and stress concentration of the material will increase the stress corrosion cracking sensitivity of stainless steel.

The research shows that in the simulated PWR primary water environment, the SCC growth rate (CGR) increases significantly with the increase of the cold working degree of stainless steel, and the IGSCC resistance of stainless steel decreases. Arioka et al. studied the SCC growth behavior of cold-worked 316 stainless steel in high temperature lithium boron solution through tensile experiments. Vispār, the crack tip is a high stress zone. The vacancy defects generated during cold processing will move towards the grain boundary under the effect of stress gradient, and move along the grain boundary to the high stress zone, forming a high vacancy density in local areas, and forming holes in the front of and around the crack, The appearance of holes and high vacancy density significantly reduces the mechanical properties at the grain boundary, weakens the binding energy at the grain boundary, provides a weak position for crack growth, and greatly accelerates crack growth. Turklāt, Terachi et al. pointed out that the vacancy and dislocation generated during cold working of 304 and 316 stainless steel samples can also significantly increase the yield strength of the material, and the crack growth rate will increase accordingly. It is generally believed that the yield strength of materials σ Between y and CGR

3.1.3 Heat treatment

Stainless steel materials generally undergo heat treatment processes such as solid solution, sensitization and aging in the production process. High temperature has a great impact on the microstructure evolution and corrosion resistance of materials. Stainless steel has more than 13% Cr content under annealing conditions, showing good resistance to general corrosion and local corrosion. Tomēr, stainless steel exposed to high temperature environment precipitates and precipitates chromium-rich carbides (Cr23C6) at the grain boundary. The chromium-poor phenomenon at the grain boundary is one of the main reasons for the reduction of intergranular corrosion resistance and stress corrosion resistance of austenitic stainless steel. Proper aging treatment can alleviate the chromium-poor problem.

For the solution treatment at a certain temperature (such as 1100 ℃), with the extension of the solution time, the solution effect of solute atoms and impurity atoms in 316L stainless steel is gradually sufficient, the microhardness increases, and the grain size increases. In the initial stage of intergranular corrosion crack, the effect of different solution treatments on the corrosion rate of the sample is not obvious; In the crack growth period, the samples with longer solid solution time showed significantly better intergranular corrosion resistance. The experimental results show that the stainless steel with solution treatment of 0.5~1h at 1100 ℃ has better comprehensive properties. Compared with solution treatment, the corrosion rate and crack growth rate of sensitized stainless steel increased significantly. Obviously, the sensitization treatment is not conducive to the improvement of the SCC resistance of 316L stainless steel. The sensitization process is easy to cause chromium deficiency at the grain boundary of 304 stainless steel. The SCC sensitivity is increased, and IGSCC is more likely to occur.