Corrosion at Nuclear Power Plants
- 29 March 2018
- Posted by: Stm Coatech
- Category: Educational Articles
- By the end of 2008, 439 power plants worldwide generated 372 gigawatts of power, while 345 power plants in OECD countries accounted for 22% of their energy needs. Three Mile Island in 1979 and the nuclear plant construction, which has slowed considerably since the Chernobyl accidents in 1986, have shown an upward trend in recent years. In Japan, for example, 53 nuclear power plants are currently operating, 30 of which are BWR and 23 of which are PWRs. The total capacity of these plants meets 47 GWe and about one third of the country’s electricity needs. Construction of 3 power plants is continuing in order to meet the increasing electricity demand and preparations are made for 8 new power plants within the next 5-10 years period. In Japan, nuclear energy is seen as an effective and practical solution to reduce CO2 emissions.Nearly half of their average 30 year life span, close to the half of the power plant in service, has passed. The dismantling of these power plants and the construction of the innovations to their places are accompanied by a great financial burden. For this reason, international cooperation for research is being developed and speeding up the work of extending the life of plants, replacing parts in critical locations with new ones made from longer-lasting materials, and making new plants more durable.
One of the most important factors in determining the lifetime of power plants is the stres corrosion cracking (SCC) problem, which is caused by the presence of tensile, corrosive and corrosive materials in the vicinity of the triple material. SCC is generally studied in three stages; start, progress and break. It is not yet possible to establish a model with broad validity in the starting phase. Life prediction studies are more focused on the relatively straightforward progress phase of the study. There are many factors that influence the start of SCC. In brief, however, the oxide layer, which has a protective effect on the surface of the metal, undergoes exfoliation and / or fracture under external influences, and the exposed surface is exposed to the corrosive environment, causing wear and tear. It is quite difficult to detect this crack which has occurred until reaching a certain paint. The crack reaching the critical dye will cause breakage quickly. For this reason, identification and prediction of crack propagation under different conditions is inevitable for determining the healthy working life.
It is quite difficult to define the mechanism of the SCC, which is the common result of various materials, environmental and mechanical parameters interacting with each other, for many possible combinations of parameters. Therefore, the estimation of the mechanism is limited to certain material and environmental conditions. In order to achieve a more general pattern, simplifications are needed that cover the basic principles that function at the center. As a result, the SCC at the point reached is defined as localized and accelerated oxidation.
Investigations conducted with the aim of understanding the breaks caused by environmental effects such as stress corrosion cracking in the light reactor environment have an important place in the development of new generation materials. Various experiments on the oxidation behavior for the extension of the lifetime of the materials used and the development of more durable materials have been carried out in order to provide precise measurements of the rate at which fractures occur in alloys in the light reactor environment and to provide a comprehensive range of numerical and theoretical modeling including quantum chemical molecular dynamics and finite element methods studies are being carried out.
Stress corrosion cracking is a kind of delayed destruction process. The crack formation and propagation velocity is about 10-6 to 10-9 m / s until the fracture phase. Beginning to appear in nuclear power plants since the 1970s, the SSC is a problem that is relatively slow to function, and it is possible to antagonize with new varieties over time. Since we are dealing with light water reactor plants in this article, we will address the LWPs encountered in boiling water reactor (BWR) and pressurized reactors (PWR). Cracks in boiling water reactors are mostly found in low-carbon stainless steel (SS 316L (N)) primary cycle piping and in cylindrical trenches located in the reactor and surrounding the reactor core. In pressurized reactors, alloy 600-and 690-reactor pressure valve control bar with nickel-based iron-chromium alloys are often encountered at source junctions with steam generator inlet nozzle.
Although the mechanism of crack formation is not fully understood yet, there is a connection between surface roughness and crack initiation start time. Defects on the surface which may occur during production or assembly, grain boundaries and weld joints are points that are highly likely to crack. Residual stresses or low values of stresses occurring during operation cause the oxide layer, which is a protective feature in these points, to weaken and fracture, causing cracks to form when the metal surface is exposed to corrosion. The cracking region that begins to progress is becoming a more complex structure. In this region, the residual material properties are different from the normal material property, and the cracked solution is different from the solution flowing on the surface.
The SCC models can be examined in two classes as anodic and cathodic models. In fact, both anodic and cadodic reactions take place during corrosion, but crack progression can be associated with one of these. The most general anodic mechanism is the wear of the material from the crack tip by active dissolution. The most general cathodic mechanism is the formation of hydrogen, absorption, diffusion and consequently the relaxation of the material. However, the mechanism models have been developed for the crack propagation phase. Several mechanisms have been proposed for the progression of stress corrosion cracking in high temperature water. In these mechanisms, the decay of the protective film layer on the surface by the chemical action on the mechanical side, the oxidation step on the crack tip and the step of rebuilding the protective film are common steps.
Until recently, material internal structure and material behavior could be studied using relatively simple test and test methods. But as material technology advances and as the need for the industry increases, the interest in understanding the complex meso, micro and nano structures of materials, their physical properties and behavior at these dimensions has increased day by day. The new material characterization techniques thus developed also offer new possibilities in studying the behavior of the SCC with the progressive mechanism operating at the center of the nanometer scale.
Since all of the SCC mechanisms are related to solid state oxidation, especially on the surface for the initial phase, on defects such as grain boundaries or notches; the progress of the oxide layer in the crack tip and the characterization of the examination of the application of these new techniques are in the field of application. Techniques such as SEM, TEM, AFM, X-ray, ESCA, SIMS, Raman and Euger spectroscopy are most commonly used in SCC surveys.
Although techniques have evolved, many problems that still need to be overcome stand in the way. It is not yet possible to fully examine crack progression dynamically. Syncrotron X-ray, femtochemistry, X-ray tomography and digital image correlation are the most extreme applications in terms of dynamically investigating, but their capacities are very limited. At this stage, computational techniques are our greatest help. Nowadays, experiment and test studies are not able to be considered separately from computer models which are known as a third branch. Modeling provides great help both in interpreting the results obtained and in deciding which experimental work should be done and which should be continued. In addition, in the virtual environment, gravitational environment, different basic material properties, very low / high pressure, temperature etc. material behaviors can be examined and new ideas can be obtained under conditions that can not be realized in reality.
The area of intensification of basic research, if it will return to the SCC problem, covers a few nanometers. Recent experimental and modeling studies at this size show that the formulas for the macro dimension used in the SCC models need to be revisited. Particularly, diffusion, electrochemistry and classical mechanical formulas have to be modified with some more suitable formulas for the nanoscale.
Another difficulty arising from corrosion in nuclear power plants is the increase in activity. In nuclear power plants, the temperature of the cooling water reaches about 300oC and the pressure reaches up to 120 bara. The pressure-bearing components in contact with the coolant are made of stainless steel or nickel-based alloys. Oxygen in the water reacts with the outermost layers of these metals to form a thin oxide layer that slows down the corrosion. Small alloy elements added to the steel can enhance the protection of this layer. Corrosion products are released from the thin metal oxide layer by the flow of the cooling water. These particles become active as they pass through the core of the reactor and are deposited on the inner surfaces of the tubes. The resulting increase in activity tends to make maintenance of nuclear power plants more costly in the long run.
Potential buildup of oxide films of activated corrosion products is a slow process that can take many years. Often the annual fuel replenishment is monitored by measures taken during the interruption.
Recently, monitoring techniques have been developed that can be used while the plant generates energy. Some of these methods have been commercialized and delivered to power plants and process industries in several countries. The laboratory-based monitoring system is used to design optimum water chemistry conditions for a plant. Similarly, the system is used to determine how changing water chemistry conditions affect a facility’s materials during shutdown and start-up operations.
As a result, stress corrosion cracking is the most important and worldwide problem in nuclear power plants. The problem of stress corrosion cracking is a complex problem involving multi-dimensional, multi-parameter and different disciplines. Since safety is much more important in nuclear power plants than in other power generation facilities, understanding the problem of stress corrosion cracking, which has a crucial role in determining the safe working life, is essential, at the very least, to ensure that the necessary measures are taken in time. However, since it is impossible to investigate all of the long term damages in different materials and geometries, different temperature, loading and working environment with classical methods, we mainly focus on the research with new experimental and computational techniques in order to understand the working mechanism, Simplified models that can be used in the field are being tried to reach.
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2. International Atomic Energy Agency, Stress Corrosion Cracking Problem in Nuclear Power Plants, Date of access: 23 March 2018,
3. Azo Materials, Corrosion in Nuclear Power Plants, Date of access: 23 March 2018, https://www.azom.com/article.aspx?ArticleID=1196