Steam Generator Of Nuclear Power Plant Research Articles

Research on flow induced vibration characteristics of heat transfer tube plugs in nuclear power plant steam generators

Research on flow induced vibration characteristics of heat transfer tube plugs in nuclear power plant steam generators

Research on magnetostrictive liquid level gauge for water level measurement of steam generator

The accurate measurement of the water level on the secondary side of the steam generator (SG) plays a crucial role in the safe and stable operation of the primary and secondary circuits of the nuclear power plant. In order to study the feasibility of applying the magnetostrictive liquid level gauge to the water level measurement of SG, the measurement results of the magnetostrictive liquid level gauge under the steady and transient conditions were obtained and compared with the traditional differential pressure liquid level gauge. The results indicated that the magnetostrictive liquid level gauge had good measurement accuracy under the cold-steady condition, thermal-steady condition and transient pressurization condition. The design parameters of float and operating parameters were necessary to correct the measurement results of magnetostrictive water level gauge. The measurement results under the transient depressurization condition were affected by the fluid movement in the container, resulting in a decrease in measurement accuracy. The results indicated that the magnetostrictive liquid level gauge had the potential to be used for water level measurement of SG in nuclear power plant.

Comparison of thermal hydraulic performance between horizontal and vertical steam generators in nuclear power plants

Abstract Steam generators (SG) for Nuclear Power Plants (NPPs) are important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. Vertical steam generators are used normally in North America, Europe and South East Asia, but horizontal steam generators are currently of great importance to reduce building heights and construction costs that govern the overall cost of new nuclear power plants. The objective of the current study is to measure the performance of nuclear steam generators in NPPs through comparative assessments of vertical and horizontal configurations. For this purpose, the vertical U-tube steam generators of a typical 4-loop Pressurized Water Reactor (PWR) was replaced by a VVER-1000 horizontal steam generators to evaluate the horizontal SG performance, at the same plant conditions and characteristics, in comparison with the vertical SG. The thermal hydraulic system code RELAP5 was used in the current work to perform the required analysis, during steady and transient scenarios.

Noise resistant steam generator water level reconstruction for nuclear power plant based on deep residual shrinkage network

Nuclear power plant (NPP) steam generator (SG) water level has strong nonlinear pattern and lagging characteristic under control system and manual adjustment. SG water level signal accidental missing and frequent abnormity may lead to incorrect operational intervention or even emergency shutdown. Deep learning SG water level signal reconstruction method trained with historical sensor data can provide supportive information for operators’ decision-making to decrease the financial cost of maintenance and improve the safety of NPP assets. However, the sensor data transferred by electronic devices can be mixed up with noise in transients or complex conditions. This study proposed a noise resistant SG water level signal reconstruction method based on deep residual shrinkage network (DRSN). The dataset in this study is collected from a power-down process collected from a NPP in China. Results from the experiments under different levels of noise are conducted to illustrate the efficacy and robustness of the proposed approaches compared with other deep learning signal reconstruction methods.

A hybrid sliding mode controller approach for level control in the nuclear power plant steam generators

This work presents a hybrid sliding mode controller approach for level control in the nuclear power plant steam generators. In a nuclear power plant, the steam generator is one of the essential pieces of equipment. Therefore, this paper aims at a robust hybrid scheme that merges internal model control concepts, sliding mode control methodology, and gain scheduling using Takagi–Sugeno multimodel fuzzy systems. Since the process presents integrating and inverse response with dead time and a highly dependent response associated with the operating power variability, this work considers process identification like an optimization problem. Hence, parallel processing algorithms such as particle swarm optimization are used. The performance achieved with the new proposal is suitable for set-point tracking and disturbance rejection.

Hybrid resampling scheme for particle filter‐based inversion

A novel online hybrid resampling (HR) scheme based on the combination of residual and multinomial resampling schemes is proposed. It can be implemented within the framework of the sequential Monte Carlo-based inversion algorithm, also known as particle filter (PF). Based upon the degeneracy of each particle, the choice of best resampling scheme among both candidates is made at each instant iteratively. Consequently, the inversion performance of PF improves by reducing the computational complexity of the resampling scheme. The proposed online HR scheme is incorporated here within the framework of the PF-based inversion scheme once applied on nuclear power plant steam generator non-destructive testing measurement data. The promising results showcase the efficacy of the technique.

Human reliability analysis for operators in the digital main control rooms of nuclear power plants

ABSTRACT As the main control room of nuclear power plants (NPPs) has been gradually digitized, new human reliability problems may emerge because of a series of new changes in the cognitive processes, behavioral patterns, and error mechanisms of operators. Aiming to address this situation, this paper proposes a method as guidance for human reliability analysis (HRA) of different cognitive Stages. This method first constructs the influencing factors of three cognitive processes, including monitoring, decision-making, and execution of actions, and then evaluates the weights of these influencing factors through an analytic hierarchy process (AHP). In this study, the parameters used in the proposed HRA method were determined by analyzing the test data obtained from a simulation model, and the results demonstrated the rationality and feasibility of the proposed method. A case example using this HRA method was given in which the human error probabilities at three stages in a nuclear power plant (NPP) steam generator tube ruptures (SGTR) accident were obtained. In summary, the proposed method is a simple and feasible HRA tool that can be applied in digital NPP main control rooms (MCRs).

Enhanced symmetrical split ring resonator for metallic surface crack detection

An enhanced sensor based on symmetrical split ring resonator (SSRR) functioning at microwave frequencies has been proposed in order to detect and characterize the metal crack of the materials. This sensor is based on perturbation theory, in which the dielectric properties of the material affect the quality factor and resonance frequency of the microwave resonator. Conventionally, coaxial cavity, waveguide, dielectric resonator techniques have been used for characterizing materials. However, these techniques are often large, and expensive to build, which restricts their use in many important applications. Thus, the enhanced bio-sensing technique presents advantages such as high measurement sensitivity with the capability of suppressing undesired harmonic spurious and permits potentially metal crack material detection. Hence, using a High Frequency Structure Simulator (HFSS) software, the enhanced sensor is modeled and the reflection S11 is performed for testing the aluminum metal with crack and without crack at the frequency range of 100 MHz to 3GHz. Variation of crack width and depth has been investigated and the most obvious finding emerged from this study is that the ability of detecting a minimum of sub-millimeter crack width and depth which is a round 10 𝛍m width or depth where the minimum shift of reflected frequency is recorded at 6.2 MHz and 3 MHz for crack width and depth respectively. The enhanced SSRR provides high capability of detecting small crack defection by utilizing the interaction between coupled gap resonators and it is useful for various applications such as aircraft fuselages, nuclear power plant steam generator tubing, and steel bridges and for others that can be compromised by metal fatigue.

Condition-based probabilistic safety assessment for maintenance decision making regarding a nuclear power plant steam generator undergoing multiple degradation mechanisms

Condition-Based Probabilistic Safety Assessment (CB-PSA) makes use of inspections and monitoring information on Systems, Structures, and Components (SSCs) to update risk quantities. In this paper, we show the benefits of exploiting the condition-based estimates for taking maintenance decisions on a SSC undergoing multiple degradation mechanisms. To develop the method, we make reference to a spontaneous Steam Generator Tube Rupture (SGTR) Accident Scenario in a Nuclear Power Plant (NPP). The SG is susceptible to multiple degradation mechanisms, i.e., Stress Corrosion Cracking (SCC) and pitting. Tube plugging and Water Lancing and Chemical Cleaning (WL-CC) can be performed, before leading to a SGTR accident. Decisions must be taken on the maintenance strategy to perform at each inspection cycle. Results of a case study regarding SGTR show that the decisions based on the risk estimates provided by a CB-PSA approach allow controlling the SGTR risk at minimum maintenance cost.

Crack Tip Stress Analysis of the Steam Generator Dissimilar Steel Welded Joint under Residual Stress Field

An typical mode of a structural integrity failure in dissimilar steel welded joints. This paper aims at studying crack tip stress of a steam generator dissimilar welded joint under residual stress field with the method of interaction integral and XFEM. Firstly, the corresponding weak form is obtained where the initial stress field is involved, which is the key step for the XFEM. Then, the interaction integral is applying to calculate the stress intensity factor. In addition, two simple benchmark problems are simulated in order to verify the precision of this numerical method. Finally, this numerical method is applying to calculate the crack tip SIF of the addressed problem. This study finds that the stress intensity factor increases firstly then decreases with the deepening of the crack. The main preponderance of this method concerns avoiding mesh update by take advantage of XFEM when simulating crack propagation, which could avoid double counting. In addition, our obtained results will contribute to the safe assessment of the nuclear power plant steam generator.

Sensitivity analysis of thermal-hydraulic parameters to study the corrosion intensity in nuclear power plant steam generators

Abstract The failure of steam generators (SGs) due to corrosion is one of the most important problems in power plants. Impurities usually accumulate in the hot sides of SG and form deposits on the SG surfaces. In this paper, the sensitivity analysis of the accumulation of water impurities in the heat exchangers of nuclear power plants is presented. The convection-diffusion equation of the liquid phase on the heated surfaces is derived and then solved by the finite volume method. Also, the effects of the thermal–hydraulic parameters in the form of dimensionless numbers, such as Pe q , Pe u , k p (relative solubility of impurity between the steam and water) on the impurities concentration are studied.

Optimization of the Double Loop Electrochemical Potentiokinetic Reactivation Method for Detecting Sensitization of Nickel Alloy 690

Nickel alloy 690 (UNS N06690) is one of the current choices for nuclear power plant steam generator tubing. The severity of certain stress corrosion cracking submodes in Ni-Cr-Fe alloys, such as Alloy 690, may be a function of the alloy sensitization condition. Sensitization results from chromium carbide precipitation at or near grain boundaries when the alloy is exposed to temperatures from 500°C to 800°C, which causes local chromium depletion in the adjacent matrix. The objective of this work was to study the parameters that control the double loop electrochemical potentiokinetic reactivation (DL-EPR) method and to develop an optimized test routine for Alloy 690. Alloy 690 specimens were tested under different thermal treatments, tailored to obtain a variety of concentrations of chromium in solid solution near the grain boundary (GB) region. The DL-EPR method was applied to Alloy 690 in solutions with different concentrations of sulfuric acid and potassium thiocyanate, at 30°C and 50°C. The optimal condition for detecting sensitization of Alloy 690 was 0.5 M H2SO4 + 0.001 M KSCN, at 30°C. Optical micrographs confirmed an intergranular type of attack under thermally aged (sensitized) conditions. The ratio of the reactivation to the activation peak intensities (Ir/Ia) obtained in the DL-EPR test agreed well with mass loss obtained in boiling nitric acid solution with Cr(VI) additions (modified Huey test). Chromium depletion profiles near the GB of thermally treated specimens under different conditions were measured with the energy dispersive x-ray spectroscopy technique, performed on a transmission electron microscope. A model available in the literature was fitted to those results. The Ir/Ia ratio for Alloy 690, obtained when using the optimal DL-EPR testing conditions, correlated well with the Cr depletion zone width and depth near the GB.

Experimental Study of the Method for Resistance Heat Treatment of Tubes for Steam Generator

The main reason for appearance and development of corrosion damage of the nuclear power plant (NPP) steam generator heat exchanger pipes is the process of stress corrosion cracking of metal under the influence of the residual tensile stresses. The method for heat treatment and straightening of the heat exchanger pipes using distension is proposed. The approbation of the process in a production environment is carried out. The construction and operation of the experimental installation are described. The experimental results show that the proposed method provides mechanical properties, grain size and resistance to intergranular corrosion of metal pipes in accordance with the requirements of Standard Specification 14-3P-197-2001 (Seamless pipe of corrosion-resistant steels with improved surface condition). In pipes, heat-treated by the proposed method, the level of residual tensile stresses and the curvature are one order less than in the tubes, heat treated by the technology which is usually used.

Water chemistry of the secondary circuit at a nuclear power station with a VVER power reactor

Results of implementation of the secondary circuit organic amine water chemistry at Russian nuclear power plant (NPP) with VVER-1000 reactors are presented. The requirements for improving the reliability, safety, and efficiency of NPPs and for prolonging the service life of main equipment items necessitate the implementation of new technologies, such as new water chemistries. Data are analyzed on the chemical control of power unit coolant for quality after the changeover to operation with the feed of higher amines, such as morpholine and ethanolamine. Power units having equipment containing copper alloy components were converted from the all-volatile water chemistry to the ethanolamine or morpholine water chemistry with no increase in pH of the steam generator feedwater. This enables the iron content in the steam generator feedwater to be decreased from 6–12 to 2.0–2.5 μg/dm3. It is demonstrated that pH of high-temperature water is among the basic factors controlling erosion and corrosion wear of the piping and the ingress of corrosion products into NPP steam generators. For NPP power units having equipment whose construction material does not include copper alloys, the water chemistries with elevated pH of the secondary coolant are adopted. Stable dosing of correction chemicals at these power units maintains рН25 of 9.5 to 9.7 in the steam generator feedwater with a maximum iron content of 2 μg/dm3 in the steam generator feedwater.

Effects of surface nanocrystallization on the corrosion behaviors of 316L and alloy 690

Surface mechanical attrition treatments (SMATs) were used to prepare nanostructured surface layers on alloys used in nuclear power plant steam generators (SGs). The effects of surface nanocrystallization on alloy corrosion behavior at room temperature and at 300°C in a simulated SG environment were studied. At room temperature, the polarization curves indicated that with increasing SMAT duration, the corrosion potential of the samples shifted negatively from smaller to larger values, and alloy active dissolution rate and passive current density both increased. Nitriding treatment was used to improve the corrosion resistance. Compared with corrosion behavior observed at room temperature, corrosion resistance in the simulated SG condition was highly enhanced because the nano-sized-grain layer formed via SMAT provided a higher density of nucleation sites for the formation of a passive film and diffusion paths for Cr, leading to the rapid formation of a dense protective oxide layer.

Application of Numerical Simulation on the Manufacturing Process Qualification of Nuclear Power Plant Steam Generator Tube-sheet

Nuclear power plant key parts-steam generator (SG) tube sheet forging operate in harsh environment, so it has high requirements for internal quality. According to RCC-M specification requirements (take CPR1000 reactor type for example), the manufacturer should carry out workshop qualification and part process qualification before delivery product manufacturing. It is commonly used that centre compaction process and conical plate upsetting process regard as two kinds of large forging upsetting ways. The numerical analogue simulation has been carried out on nuclear power large tube sheet forgings using France Forge simulation software. The results show that: when the same pressure size, 3-direction compressive stress in the core of the billet always exist in the process of upsetting using centre pressing process, and the strain create in the central of the billet, but stagnant zone exist on the bottom and end plane; tensile stress in the core of the billet always exist in the process of upsetting using conical plate upsetting process, it will force the billet that contact with conical plate to deform, so eliminate stagnant zone, but effective deformation don’t produce in the central of the billet. The conical plate continue press for conical plate upsetting process, tensile stress in the core of the billet disappear, and then gradually change into pressure stress, then tensile stress is generated at the bottom, finally tensile stress disappear and 3-direction compressive stress exist in the core of the billet.

Crack initiation and propagation under fretting fatigue of inconel 600 alloy

The purpose of the study is to experimentally investigate the fatigue crack initiation and propagation characteristics of nuclear power plant steam generator tube materials, Inconel 600 alloy, under fretting fatigue conditions. In addition, this study aims to enhance the overall reliability of nuclear power plant equipments by developing fretting fatigue life and integrity evaluation methods. In order to achieve these purposes, a fretting fatigue experiment apparatus is developed and fatigue crack growth experiments are conducted to evaluate the characteristics of Inconel 600 alloy under various stress conditions. Finally, microstructures are observed in accordance with fretting fatigue and contact pressure conditions and fretting fatigue crack initiation mechanism is examined.

Mechanical calculation and evaluation of supports for nuclear power plant steam generator

Mechanical calculation and evaluation of supports for nuclear power plant steam generator

Evaluation of the Effectiveness of a New Technology for Extraction of Insoluble Impurities from Nuclear Power Plant Steam Generators with Purge Water

An experimental technology for the removal of insoluble impurities from a horizontal steam generator with purge water during planned shutdowns of the power generating unit is improved through a more representative determination of the concentration of impurities in the purge water ahead of the water cleanup facility and a more precise effective time for the duration of the purge process. Tests with the improved technique at power generating unit No. 1 of the Rostov Nuclear Power Plant show that the efficiency with which insoluble impurities are removed from the steam generator volume was more than two orders of magnitude greater than under the standard regulations.

Deposit models for tube support plate flow blockage in Steam Generators

Corrosion product deposits in the secondary side of nuclear power plant Steam Generators may result in Tube Support Plate flow blockage, and tube fouling. In order to simulate those two phenomena in the whole Steam Generator, a solid deposit growth model has been developed by the EDF R&D Division. This model is implemented in the frame of THYC, which is the EDF's reference code for the modeling of two-phase thermal-hydraulic phenomena at the subchannel scale. A new deposit process, based on Tube Support Plate flow blockage studies, has been developed and implemented in the model, and is presented in this work. It can be defined by two main steps: particle deposition, and strengthening process called “flashing” due to soluble species precipitation in the pores of the particle deposit. The relevance of this process is tested by comparing the simulation results to the actual levels of flow blockage observed in some nuclear plants. Two dominant trends are showed in this work: the flow blockage is more important on the hot leg than on the cold leg and at the top than at the bottom of the Steam Generators. Moreover the flow blockages at the upper Tube Support Plate have the special feature to be more important at the periphery than at the center. The “flashing” phenomenon allows one to underline the magnetite solubility dependence, so the pH dependence, of flow blockage phenomenon. A pH elevation of the secondary circuit seems to be a interesting remedy which is currently considered on EDF fleet.