Feedback control of a large steam generator

Feedback control of a large counter current heat exchanger employed as a once-through steam generator for a breeder reactor power plant is presented. The mathematical model of the exchanger consists of a set of nonlinear partial differential equations with boundary conditions specified at three space points. The system structure does not allow the control signals to be distributed in space; they are applied only at the boundary points.The objective of the investigation is to choose a practical implementable time-invariant feedback controller which would give satisfactory regulation of the system. Although the emphasis is placed on a specific once-through steam generator system, the techniques used here are equally applicable to other heat exchanger systems.     

Theoretical study on a multivariate feedback control of a sodium-heated steam generator

This paper applies the connection of a multivariate feedback controller with a state estimator to a 1-MW sodium-heated steam generator for LMFBR theoretically, to obtain a control strategy which emphasizes, from the view point of safety and availability of the FBR plant, that a superheat of 30°C should be required for the evaporator steam. This involves a trial to study the feasibility for the estimation of such an inaccessible variable as the dry-out location of tubes and utilize the state estimate to design a feedback controller of steam generators. The Kalman filter tested was found to generate reasonable estimates of the transient process variables of the steam generator and can provide a major advantage of regulating steam condition of the system even in the presence of contamination by a rather high level of measurement noise in the view point of economic uses of micro- and/or minicomputers.     

Dynamic simulation of a steam generator by neural networks

Numerical simulation by computers of the dynamic evolution of complex systems and components is a fundamental phase of any modern engineering design activity. This is of particular importance for risk-based design projects which require that the system behavior be analyzed under several and often extreme conditions. The traditional methods of simulation typically entail long, iterative, processes which lead to large simulation times, often exceeding the transients real time. Artificial neural networks (ANNs) may be exploited in this context, their advantages residing mainly in the speed of computation, in the capability of generalizing from few examples, in the robustness to noisy and partially incomplete data and in the capability of performing empirical input–output mapping without complete knowledge of the underlying physics. In this paper we present a novel approach to dynamic simulation by ANNs based on a superposition scheme in which a set of networks are individually trained, each one to respond to a different input forcing function. The dynamic simulation of a steam generator is considered as an example to show the potentialities of this tool and to point out the difficulties and crucial issues which typically arise when attempting to establish an efficient neural network simulator. The structure of the networks system is such to feedback, at each time step, a portion of the past evolution of the transient and this allows a good reproduction of also non-linear dynamic behaviors. A nice characteristic of the approach is that the modularization of the training reduces substantially its burden and gives this neural simulation tool a nice feature of transportability.     

Exploratory research on creep and fatigue properties of 9Cr-steels for the steam generator of an FBR

Research on the applicability of 9Cr-steels to the steam generator of the demonstration fast breeder reactor was performed by the Subcommittee of the Japan Welding Engineering Society as a four-year program from 1985. In this program, exploratory tests, which included tensile, creep rupture and low-cycle fatigue tests, were conducted on three kinds of 9Cr-steels (Mod.9Cr-1Mo, 9Cr-1Mo-V-Nb, and 9Cr-2Mo) and their weldments. This paper describes the summary of results obtained in this program. Among the tested 9Cr-steels, Mod.9Cr-1Mo steel shows the best creep rupture strength and its weldment indicates almost the same level of creep rupture strength and the base metal at 500 and 550°C. The low-cycle fatigue properties of Mod.9Cr-1Mo steel is also discussed from its relation to the tensile properties.     

Droplets deposition in steam piping connecting steam generator and steam turbine in nuclear plant

A numerical investigation is carried out for turbulent droplet-laden flow of saturated steam produced in a steam generator (SG) that feeds steam turbine (ST) through a long and multi-bend steam piping. The main purpose of the study is to analyze deposition of droplets that form a wall film in the piping system. Commercial CFD code StarCD is used for the solution of turbulent flow field of droplet-laden steam. Turbulence is treated using k–ω model of turbulence. Wall film formation is solved by additional conservation equations. Two tasks were performed: parametric study of the deposition in a 90° elbow positioned with different orientation and the deposition in a more complex piping system. This system starts with outlets from steam generator with five mouthpieces leading to a collector pipe and connecting the steam piping leading to a steam turbine. The steam piping consists of three straight segments of pipes and two 90° elbows in the total length 17 m. The diameter of the steam piping is 0.425 m. Results of the simulations show where droplets deposit and where a liquid separator should be placed to drain away the water film and to avoid droplets from entering the steam turbine.     

Theoretical study of flow instability of a sodium-heated steam generator

A representation of a sodium-heated steam generator was formulated to study density-wave oscillation of the flow. The mathematical model, presented (designated SGEX code), facilitates prediction of flow instability by examining movements of unstable eigenvalues of the system matrix. It was defined from the trajectory of the system on the pressure loss and inlet water velocity plane when dynamic instability occurred during the flow instability test of the 1 MW (th) sodium-heated s steam generator in OHARAI. Inception of flow instability may be attributed to a large pressure loss in the superheated region, caused by increase in sodium flow rate and inlet temperature and decrease in feedwater flow rate.     

新型蒸汽发生器泄漏监测装置的研制

本装置的工作原理是基于堆芯中子与冷却剂中轻水或重水中的氧同位素16O、18O核反应时,其放射性产物16N、19O的半衰期各不相同,由于16N和19O从堆芯通过不同的泄漏点到达探测点所经历的路程不同,因此16N和19O的活度比值也随之不同,通过所测得的不同活度比,则可以反推出U形管泄漏的位置.将研制的监测装置在本院游泳池反应堆一回路水取样室进行了测量.结果表明,本装置性能良好,达到了设计的要求.     

直流蒸汽发生器的广义预测自校正控制仿真研究

为了进一步提高直流蒸汽发生器的动态控制性能,在阐述了直流蒸汽发生器的数学模型的基础上,提出将广义预测自校正控制算法应用于直流蒸汽发生器主要参数的控制中,包括控制结构和控制器设计.仿真结果显示当蒸汽发生器经历较大的变工况负荷扰动时,广义预测自校正控制律比经典PID控制律下的蒸汽发生器主要参数变化响应要快10 s左右,而经典PID控制律下的主要参数值要比广义预测自校正控制律产生较大的超调.表明所采用的广义预测自校正控制算法能够较好的控制直流蒸汽发生器主要参数的输出,可以获得较好的控制效果.     

蒸汽发生器工作过程建模及仿真分析

基于分布参数热工对象的集总参数化动力学模型,对自然循环蒸汽发生器进行了控制体划分并建立了数学模型,并用MATLAB语言和SIMULINK仿真软件对其进行了仿真,文章采用了Runge-Kutta (4,5)求解器,得到不同功率装置运行时,一次侧,二次侧,以及管束的温度分布,并得到一回路给水扰动时,传热量以及冷却剂出口焓的响应曲线.     

Effects of power level on thermal-hydraulic characteristics of steam generator

The thermohydraulics of the steam generator secondary is analyzed using Fluent's porous media model coupled to a two-phase flow mixture model. The resistances introduced by the tubes, supports, downcomers and separators are calculated as the source terms of the momentum equation. The heat transfer from primary side fluid to the secondary is calculated three-dimensionally each iteration and is supplied as a heat source on the secondary flow field calculation. The effects of the power level on thermal hydraulic characteristics in the steam generator are also investigated. Flow-induced vibration damage increased in severity with increasing power level.     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

Vibration analysis of heat exchanger and steam generator designs

A thorough flow-induced vibration analysis of nuclear components such as heat exchangers and steam generators is essential at the design stage to ensure good performance and reliability. This paper presents our approach and techniques in this respect. In a steam generator, for example, the flow may be liquid or two-phase. In general, parallel and cross-flow exist in the tube bundles of heat exchange components. In cross-flow three basic vibration excitation mechanisms are considered, namely fluidelastic instability, periodic wake shedding resonance, and forced response to random flow turbulence. The latter may need to be considered in parallel flow. These vibration excitation mechanisms and the dynamics of multispan tubes are formulated in a computer model which is used to predict the vibration response of the tubes. The computer model and the parameters required to formulate the vibration excitation mechanism are discussed. Examples of vibration analysis of steam generators and heat exchangers are outlined. It is concluded that most flow-induced vibration problems may be avoided by proper analysis at the design stage.