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Nuclear steam generators (NSGs) are key components of nuclear power plants (NPPs) since their reliability affects the overall plant performance. Corrosion control of steam generators under operation and during plant outages is attained by proper design, adequate material selection and control of water chemistry in the entire secondary cooling circu...
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... superheated steam generators have straight tubes which carry the primary water from top to bottom ( Figure 2). Typically, the primary coolant enters the steam generator at 316-327 °C and leaves at 291-293 °C. ...
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... concentration in feedwater may be reduced down to 7 μg/kg by mechanical deaeration. Further oxygen removal is accomplished by adding a scavenger, which is generally hydrazine (N 2 H 4 ). Hydrazine reacts with oxygen according to Eq. (1) and it also thermally decomposes into ammonia according to Eq. (2). ...
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... EDF prohibits the use of a nitrogen blanket due to safety hazards (EPRI 2005;Long et al. 2002). Figure 12 shows a potential vs. pH diagram for various half cells potentially involved in the wet lay-up of steam generators and BOP, at 25 °C. The pH range of interest for this application is roughly between 9 and 11. ...
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... reduction of N 2 H 4 is quite slow at room temperature and open circuit potentials of metals are well above the N 2 /N 2 H 4 half-cell potential. Figure 12 shows open circuit potentials of carbon steel A106C and alloys 600, 690 and 800 in simulating lay-up solutions. Passive alloys show significant potential ennoblement with regard to the actively corroding carbon steel ( Rodríguez and Kappes n.d.) EDF carried out a study on the appropriate conditions for wet lay-up of carbon steel components by varying the gas phase and the hydrazine concentration. ...
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... chloride, another impurity usually present in the secondary circuit is sulphate. In a reducing environment, sulphate may be reduced to aggressive lower-valence sulphur species, such as thiosulphate or sulphide (Figure 12). Thiosulphate has shown to be particularly detrimental for tubing alloys in conjunction with chloride ( Xia et al. 2013). ...
Citations
... This was observed in alloy 600 (15 at% Cr) and justified its replacement by alloy 690 (30 at% Cr) [9]. Consequently, the latter is used in many manufactured components of both PWR and BWR, such as steam generators and vessel head penetration nozzles [10,11]. Yet, in as much as healing is associated with the consumption of chromium, as of today, it is not clear whether alloy 690 becomes prone to cracking after a very long (approx. ...
... Costs to avoid scaling and anti-scaling treatment can be as high as USD 0.93 million per year or 0.88 % of revenue for a 550 MW base plant [11]. In order to effectively control the formation of sludge and scale in RCS, it is necessary to understand the processes of the behavior of DP in the RCS of power plants [12][13][14]. Common sources of DP in RCS are impurities in makeup water, which is used to replenish losses in the system; corrosion and erosion of structural materials; biological fouling; therefore, to determine the sources of entry, it is necessary to establish the chemical composition of DP. ...
The object of this study is the processes of formation and changes of dispersed particles in fresh, make-up, cooling, and return water in open recirculating cooling systems (RCS) with an assessment of the influence of suspended substances in discharge waters on the aquatic ecosystem. The study was carried out on the example of the Rivne Nuclear Power Plant (RNPP) and the Styr River. Dispersed particles (DPs) pose technological obstacles in the RCS of power plants, and their content in discharge waters determines the ecological quality of water bodies. This paper describes the results of studying the formation and changes of DP in raw, make-up, cooling, and return waters of RNPP RCS with an assessment of the impact of suspended substances in discharge waters on the aquatic ecosystem of the Styr River. It was found that the formed dispersed particles after water treatment by liming contain DP consisting of calcium carbonate and have a size of 10–30 μm. As a result of agglomeration of DP in RCS, they aggregate to 120–150 μm, and due to low sedimentation resistance (sedimentation time 0.97 h), they settle in RCS. As a result of the deposition of DP in RCS, their significant decrease in return water (min–max=7.31–16.12 mg/dm3) is observed, despite the increase in their content in make-up water after water treatment (min–max=10.22–49.46 mg/dm3). According to the ecological classification, according to the content of suspended substances, the water of the Styr River in the zone of influence of RNPP discharges belongs to the II class, category 2, which characterizes the quality of the water as "very good" in terms of its state, and "clean" in terms of its degree of purity. It was concluded that the content of suspended solids does not exceed the established maximum permissible concentration (25 mg/dm3), the increase in the concentration of suspended solids does not exceed the established ecological standard of 0.25 mg/dm3 and does not have a negative impact on surface water. The results of the research could be used for other power plants equipped with an open RCS
... These extensive mechanisms affect both the primary and secondary sides, with secondary-side degradation being a major concern. Additionally, numerous SG replacements are required, incurring costs related to repair work, radiation exposure for personnel, and power loss [20,21] Several explanations regarding degradation mechanisms have been published [22][23][24]. They are summarized in Fig 2, which identifies the degradation locations for PWR SG. ...
In Pressurized Water Reactor (PWR) Nuclear Power Plants (NPPs), the steam generator is crucial for transferring heat from the primary to secondary cooling systems, vital for steam production to drive turbines, and central to nuclear power safety. This study explores recent research on multi-axial loading, structural integrity, and material durability in PWR steam generators, shedding light on key factors affecting these systems. Common corrosion-related degradation in steam generators often arises from design, material, and water chemistry factors. However, the shift to All Volatile Treatment (AVT), the development of advanced material alloys, and enhanced water quality control in primary and secondary systems have significantly reduced instances of steam generator degradation. These findings promise to enhance the reliability and safety of steam generators in future nuclear applications.
... SCC is one of the most common cause of structural and component failures by the source of harmful corrosion in the nuclear, oil & gas, and petrochemical, as effects unexpected service failures and occurs significant economic and environmental implications [15][16][17][18]. Especially within nuclear reactor components, such as pipes, heat exchangers/coolant system, steam generators and other internals, those composed of ASS can exhibit vulnerability to CI-SCC and owing to the presence of chloride ions in the coolant and the prevailing operating conditions [19,20]. ...
... The types SS304 and SS316 are extensively utilized for pressure boundary piping in BWR (Boiling Water Reactors) and for primary circuits in PWR (Pressurized Water Reactors). Furthermore, SS316 materials with add-on chemical elements such as titanium (Ti) and slightly changes in Si and P are employed for fuel cladding tubes in light water reactors such as sodium-cooled fast reactors [20,61]. They have such excellent mechanical properties such as Tensile strength (691 MPa), Yield strength (410 MPa) and Elongation 66.5 % [37]. ...
Austenitic stainless steels (ASS) are extensively employed in various sectors such as nuclear, power, petrochemical, oil and gas because of their excellent structural strength and resistance to corrosion. SS304 and SS316 are the predominant choices for piping, pressure vessels, heat exchangers, nuclear reactor core components and support structures, but they are susceptible to stress corrosion cracking (SCC) in chloride-rich environments. Over the course of several decades, extensive research efforts have been directed towards evaluating SCC using diverse methodologies and models, albeit some uncertainties persist regarding the precise progression of cracks. This review paper focuses on the application of Acoustic Emission Technique (AET) for assessing SCC damage mechanism by monitoring the dynamic acoustic emissions or inelastic stress waves generated during the initiation and propagation of cracks. AET serves as a valuable non-destructive technique (NDT) for in-service evaluation of the structural integrity within operational conditions and early detection of critical flaws. By leveraging the time domain and time-frequency domain techniques, various Acoustic Emission (AE) parameters can be characterized and correlated with the multi-stage crack damage phenomena. Further theories of the SCC mechanisms are elucidated, with a focus on both the dissolution-based and cleavage-based damage models. Through the comprehensive insights provided here, this review stands to contribute to an enhanced understanding of SCC damage in stainless steels and the potential AET application in nuclear industry.
... If the radioactivity of the outgoing steam is higher than the regulatory limit, the NPP must be shut down immediately, resulting in potential power outages. A review of twenty pressurized water reactors in the Unites States shows that the risk associated with SG tube failure can be as high as 75% of the total plant risk [3,4]. This necessitates the need for corrosion control of the SG tubes, which can be achieved through a combination of materials selection and water chemistry control [5,6]. ...
... The material used in the 1960's in the CANDU SGs was alloy 400 (UNS N04400) with good corrosion properties and reasonable cost. Unfortunately, alloy 400 is very sensitive to the presence of oxygen, even at very low concentrations [3]. Alloy 800 (UNS N08800), an alternative for alloy 400, has been used for SG tubing since the early 1970's [1,7]. ...
... Table 1 shows the secondary chemistry guidelines for a 600 MWe CANDU reactor with Cu containing materials [4]. As mentioned, the temperature and pressure at the secondary side of SG tubes are 260 ℃ and 4.7 MPa, respectively [1][2][3]; therefore, E-pH diagrams were constructed at 100, 200, and 260 ℃ under 4.7 MPa pressure with different concentrations of ammonia or acetate ion. ...
Nuclear power plant steam generator (SG) tubes contain nickel as the main alloying element and there is concern about their corrosion. We optimized models for calculating the high-temperature and high-pressure thermodynamic properties of common nuclear alloy elements and then, calculated the E-pH diagrams for nickel in different concentrations of ammonia or acetate ion between 100 to 260 ℃ at 4.7 MPa. This information is used to predict the corrosion behaviour for nickel in the secondary circuit conditions of CANDU SGs. Based on the results, nickel ammonia and nickel acetate complexes are not stable at low concentrations of ammonia or acetate ion. However, increasing the ammonia or acetate ion concentration resulted in the predominance of Ni(NH3)n2+ or Ni(Ac)n(2−n) at the expense of Ni2+ and NiO, indicating a higher risk of nickel corrosion. Calculations showed that under normal operating conditions with [NH3]tot=5×10−5 and [CH3COO−]tot=10−8 m at 260 ℃ and 4.7 MPa, nickel will be passivated as NiO, preventing the rapid degradation of nickel-based alloys. However, in the presence of a crevice that allows the acetate ion to concentrate, nickel would dissolve in the form of the Ni2+ ion, endangering safe operation of CANDU SGs.
... Using octadecylamine(ODA) to protect carbon steel components exposed to corrosive environment has been proved to be an effective method [3]- [4][5] [6]. For example, Mao et al. [7] used electrochemical impedance spectroscopy to study the inhibition effect of ODA on type 316SS steel immersed in 348 K, acetic acid solution with a pH of 4.2 and found that the charge transfer resistance was 640945 Ωcm2 with ODA concentration at 0.01 M which was 19973 Ωcm2 when without ODA. ...
Octadecylamine (ODA), a well-known film-forming amine, recently finds increasing application in nuclear pressurized water reactors to protect carbon steel components in the secondary circuit against corrosion. To clarify the inhibition mechanism, molecular dynamics simulations were employed to study the diffusion characteristics of three commonly encountered species, H2O, O2, and Cl-, in the formed inhibitor film. The simulation results revealed that the formed film mitigates corrosion by effectively impeding the diffusion of corrosion species in the film. For various species, the film exhibits markedly different inhibition abilities.
... Pitting corrosion can appear during the lay-up of SGs when S 2 O 3 2-, chloride (Cl -), oxygen and other oxidants are combined and accumulated in the secondary circuit [10]. Here, S 2 O 3 2can come from the reduction of sulfate (SO 4 2-) impurities in the presence of hydrazine (N 2 H 4 ) used in the all-volatile water treatment for SG [10,11]. ...
Alloys 600 and 800 exhibited stable pitting corrosion in 1 M NaCl plus thiosulfate deaerated solutions at the open circuit potential (OCP) and room temperature. Alloy 690 exhibited only metastable pitting. Stable pitting, stable passivity or metastable pitting were identified by observing features in the OCP vs. time transients and by analysis of electrochemical impedance spectroscopy measurements. Post-mortem scanning electron and optical microscopy observations confirmed these findings. Stable pits nucleated adjacent to Ti-rich inclusions, and reached diameters above 100 μm after 50 h of OCP immersion. A prepassivation treatment inhibited this mechanism of pitting corrosion.
... If a strong SG can directly heat the inlet steam of the gasifier with the helium from the HTR, the temperature of the inlet steam will be around 570 °C which can increase the plant efficiency by around 2.5%. However, the corrosion of the SG tubes and the deposits on them cannot be ignored even in the highly purified conditions of the nuclear reactor system [34,35]. It is not economically feasible to make the water for biomass gasification highly purified. ...
Biomass gasification to produce burnable gas now attracts an increasing interest for production flexibility in the renewable energy system. However, the biomass gasification technology using dual fluidized bed which is most suitable for burnable gas production still encounters problems of low production efficiency and high production cost. Here, we proposed a large-scale biomass gasification system to combine dual fluidized bed and high-temperature gas-cooled reactor (HTR) for co-production of hydrogen and synthetic natural gas (SNG). The design of high-temperature gas-cooled reactor biomass gasification (HTR-BiGas) consists of one steam supply module to heat inlet steam of the gasifier by HTR and ten biomass gasification modules to co-produce 2000 MWth hydrogen and SNG by gasifying the unpretreated biomass. Software for calculating the mass and energy balances of biomass gasification was developed and validated by the experiment results on the Gothenburg biomass gasification plant. The preliminary economic evaluation showed that HTR-BiGas and the other two designs, electric auxiliary heating and increasing recirculated product gas, are economically comparative with present mainstream production techniques and the imported natural gas in China. HTR-BiGas is the best, with production costs of hydrogen and SNG around 1.6 $/kg and 0.43 $/Nm3, respectively. These designs mainly benefit from proper production efficiencies with low fuel-related costs. Compared with HTR-BiGas, electric auxiliary heating is hurt by the higher electric charge and the shortcoming of increasing recirculated product gas is its lower total production. Future works to improve the efficiency and economy of HTR-BiGas and to construct related facilities are introduced.
This review paper provides a comprehensive insight about the issue of SCC in NPPs, encompassing its economic impacts, mechanisms, material considerations, testing strategies, predictive modeling, and prevention methods. The study will explore the relationship between materials and environmental variables which cause the onset and progression of SCC in different materials commonly used for piping systems and related infrastructure. SCC stands as a noteworthy concern within the nuclear power industry, resulting in expensive downtime and repair procedures, and carrying potential severe safety implications. This comprehensive review will provide insights into the mechanisms underlying SCC and development of strategies for preventing and mitigating its effects on piping systems. By understanding the factors that contribute to SCC and how it can be mitigated, industries can develop strategies to prevent SCC related failures and ensure safe and reliable operation.
