Dynamic Model Identification of Once-Through Steam Generation System Based on PEM Method

2017 ◽  
Author(s):  
He Jinliang ◽  
Tao Mo ◽  
Wang Wei ◽  
Song Feifei
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Step Response ◽  
Steam Generation ◽  
Generation System ◽  
Two Phase ◽  
Coupling System ◽  
Transfer Tube ◽  
Mechanism Modeling ◽  
Secondary Side

Once-Through Steam Generator (OTSG) has the advantages of simple structure, good static characteristics, and can produce superheated steam. In recent years it has been widely used in nuclear power system for its good maneuverability. The working fluid flows once through the heat transfer tube forced by the pressure head of the feedwater pump. The water is heated by the coolant of primary side as it flows along the heat transfer tube. After preheating, evaporating and overheating, steam of required temperature is then produced. In addition, flow instability may occur in OTSG because of the present of two-phase flow. Once-through steam generation system is a typical nonlinear multi-variable coupling system, the water/heat storage capacity of secondary side is small, and the steam pressure is very sensitive to the load fluctuation. Various disturbances such as the change of feed water flow, the change of heat transfer rate from the primary side and the change of the secondary side load, will lead to the variety of the parameters in the heating channel. A more complex and precise water control system is therefore needed. An accurate system model is very important to optimize the control strategy and improve the control quality. Obtaining the dynamic mathematical model of the once-through steam generation system is the basis for its effective control. At present, the dynamic modeling of once-through steam generation system mainly adopts mechanism modeling method. By analyzing the inherent mechanism of steam generation process and using the basic conservation equation to derive the relationship between model variables. However, due to the complexity of OTSG and the feedwater system structure and the two-phase heat transfer mechanism, the mechanism modeling is very difficult and the model is not precise enough and very complex. Many nonlinear equations are included, which makes it difficult to determine an effective numerical method for real-time simulation. System identification is based on the actual measurement of input and output information of the process. The system model can be estimated without having to study its internal mechanism. In this paper, the dynamic model of once-through steam generation system in nuclear power plant is identified. Ensuring the stability of the steam outlet pressure during the operation of the system is very important to the safety of steam turbine. Therefore a three-input, two-output coupling system is obtained by analyzing the influence factor of the once-through steam generation system. The pseudo-random sequences are used as the input signal, and the Prediction-Error Minimization (PEM) method is used to identify the system. Dynamic state space models of the system are obtained. The multi-input and multi-output (MIMO) system are identified at different power levels, and the model verification are carried out by simulate the step response output. The results show that the state space model of the once-through steam generation system identified by PEM method is of high precision. The step response of the model and the output of the actual system are in good agreement with each other. The identification scheme proposed in this paper provides a new method and idea for the modeling of nuclear power plant system. The model can provide foundation and technical support for the research of high precision and high quality control system.

Characteristic Optimization of the Double-Tube OTSG

2010 ◽  
Author(s):  
Xinyu Wei ◽  
Fuyu Zhao ◽  
Yun Tai ◽  
Chunhui Dai
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Flow Resistance ◽  
Optimization Problem ◽  
Great Influence ◽  
Optimization Method ◽  
Straight Tube ◽  
Transfer Enhancement ◽  
Transfer Region ◽  
Secondary Side

The OTSG (Once-Through Steam Generator) is usually used in the integral nuclear power equipment which requires smaller size and better effect of heat transfer. The OTSG with double-side heat transfer component is presented in this paper. The heat transfer component is composed of straight tube outside and helix tube inside. In the both sides of the helix tube, the flow is spirally, therefore, the heat transfer is enhanced. The smaller the pitch, the stronger the spirally flow, the effect of heat transfer is better, but the flow resistance is raised. Especially the increased flow resistance in the secondary side brings a great influence to the pump. The heat transfer region of the secondary fluid are divided into three regions: sub-cooled region, boiling region, and superheated region, the effects of heat transfer induced by the spirally flow vary in different regions. Thus, there is an optimization problem which is to find an optimization pitch of the inner helix tube with the best effect of heat transfer and the minimum flow resistance. Based on analyzing the effects of the pitch on heat transfer enhancement and flow resistance, the pitch is optimized by the constrained nonlinear optimization method.

Development of a Thermal Energy Storage System for Parabolic Trough Power Plants With Direct Steam Generation

2010 ◽  
Vol 132 (2) ◽  
Author(s):  
Doerte Laing ◽  
Thomas Bauer ◽  
Dorothea Lehmann ◽  
Carsten Bahl
Keyword(s):  
Heat Transfer ◽  
Power Plants ◽  
Storage System ◽  
Thermal Power ◽  
Sensible Heat ◽  
Steam Generation ◽  
Parabolic Trough ◽  
Two Phase ◽  
Solar Thermal Power ◽  
Direct Steam

For future parabolic trough plants direct steam generation in the absorber pipes is a promising option for reducing the costs of solar thermal power generation. These new solar thermal power plants require innovative storage concepts, where the two-phase heat transfer fluid poses a major challenge. A three-part storage system is proposed where a phase change material (PCM) storage will be deployed for the two-phase evaporation, while concrete storage will be used for storing sensible heat, i.e., for preheating of water and superheating of steam. A pinch analysis helps to recognize interface constraints imposed by the solar field and the power block and describes a way to dimension the latent and sensible components. Laboratory test results of a PCM test module with ∼140 kgNaNO3, applying the sandwich concept for enhancement of heat transfer, are presented, proving the expected capacity and power density. The concrete storage material for sensible heat was improved to allow the operation up to 500°C for direct steam generation. A storage system with a total storage capacity of ∼1 MWh is described, combining a PCM module and a concrete module, which will be tested in 2009 under real steam conditions around 100 bars.

Power Rising and Descending Transient for the OTSG of a Small PWR

2020 ◽  
Author(s):  
Baihui Jiang ◽  
Zhiwei Zhou ◽  
Zhaoyang Xia ◽  
Qian Sun
Keyword(s):  
Heat Transfer ◽  
Flow Rate ◽  
Two Phase Flow ◽  
Flow Instability ◽  
Flow Stability ◽  
Phase Flow ◽  
Steam Generators ◽  
Two Phase ◽  
Water Reactors ◽  
Secondary Side

Abstract As key heat transfer system in small and medium size pressurized water reactors, once-through steam generators are important parts of energy exchange between primary and secondary circuits, and are very important for the design and operation of reactors. However, two-phase flow and heat transfer in once-through steam generators are very complicated. When a reactor experience power rising and descending transient, the heat removal of once-through steam generator, the flow rate, the inlet fluid temperature and outlet steam temperature will all change accordingly. Especially when a reactor is running at a low power, the flow rate of the secondary side of OTSG is extremely small and the single-phase region of the secondary side of OTSGs is also too small. The two-phase flow instability may occur, which has a serious impact on reactor operation and safety. So, a reasonable power-up and power-down transient scheme is required to ensure operational stability when starting up and shutting down a reactor. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LCC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. Scholars all over the world have carried out a large number of analysis of two-phase flow stability using RELAP5, and the results are reliable. This paper takes once through steam generators with given structural parameters as the research object, and uses RELAP5 as the calculation tool. The influencing factors of flow instability are discussed in this paper, and the operating parameters of the fluid on the primary and secondary sides are designed to satisfy the flow stability under different powers. And a set of power-up and power-down schemes for stable operation is proposed.

The Simulation Research of Noncondensable Gas to Condensation in Secondary Side Condenser of Floating Nuclear Power Plant Based on RELAP5

2018 ◽  
Author(s):  
Si-wei Yan ◽  
Chun-mei Li ◽  
Tie-bo Liang ◽  
Jing Zhao ◽  
Cheng-ming Hao ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling Water ◽  
Gas Content ◽  
Simulation Research ◽  
Cooling Water Temperature ◽  
Noncondensable Gas ◽  
Secondary Side

Similar to conventional nuclear power plant, condensate water subcooling is a common problem in secondary coolant of floating nuclear power plant, which is caused by many reasons. In this article, RELAP5 is used to simulate the phenomenon of condensate water subcooling caused by noncondensable gas. The influence of noncondensable gas to condenser pressure, subcooling temperature, heat transfer rate, terminal temperature difference, cooling water temperature rise is presented. The results obtained through this study have shown that the model with non-condensable gas in steam can simulate condensate water subcooling, and reveal the discipline of condenser heat transfer characteristics as a function of noncondensable gas content.

Study on Flow Induced Vibration Analysis and Evaluation for Heat Transfer Tube of Steam Generator

2018 ◽  
Author(s):  
Xuan Huang ◽  
Huan-Huan Qi ◽  
Feng-Chun Cai ◽  
Zhi-Peng Feng ◽  
Shuai Liu
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Vibration Analysis ◽  
Two Phase Flow ◽  
Phase Flow ◽  
Flow Induced Vibration ◽  
Two Phase ◽  
Heat Transfer Tube ◽  
Analysis And Evaluation ◽  
Transfer Tube

The heat transfer tube of steam generator is an important part of the primary loop boundary, the integrity is crucial to the safe operation of the whole reactor system; the flow induced vibration is one of the main factors leading to the failure of heat transfer tube in steam generator. Both ASME and RG1.20 have made a clear requirement for the analysis and evaluation of the flow induced vibration of steam generator. The flow induced vibration of heat transfer tube in two-phase flow is the difficult and important content in the analysis. In this paper, the finite element model of heat transfer tube is established and the modal analysis is carried out. Then in order to evaluate the influence of two-phase flow in the secondary side and support boundary constraint, the analytical results are compared with the natural frequencies of the heat transfer tube measured in the two-phase flow test. On the basis of accurate simulation of the dynamic characteristics of heat transfer tube in two-phase flow, the paper calculate the turbulent excitation response and the fluidelastic instability ratio aiming at the main mechanism causing the flow induced vibration of heat transfer tube in two-phase flow. Firstly, the modified PSD of turbulent excitation is proposed on the foundation of root mean square displacement amplitude of heat transfer tube measured in two-phase flow test. The calculation result of the amplitude of heat transfer tube with different void fraction can envelope the test result by using the modified PSD as input, and the safety margin is reasonable. Then we also verify whether the analysis conclusion of fluidelastic instability is in agreement with the test. Finally, the analytical technique is applied to the analysis of flow induced vibration of steam generator to verify the design of structure. The paper studies on flow induced vibration analysis and evaluation a heat transfer tube of steam generator in two-phase flow. The analysis program of flow induced vibration on the basis of the test results. The investigation can be used for the risk prediction and evaluation of flow induced vibration of heat transfer tube in two-phase flow, solve the technical difficulties of flow induced vibration analysis in two-phase flow, and provide the technical support for the flow induced vibration analysis of steam generator.

scholarly journals A heat transfer tube wear reliability analysis method based on first-order reliability method

2020 ◽  
Vol 7 (6) ◽  
pp. 803-815
Author(s):  
Xiao Chen ◽  
Xing He ◽  
Lichen Tang ◽  
Yuebing Li ◽  
Mingjue Zhou ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Reliability Analysis ◽  
Nuclear Power ◽  
Engineering Practice ◽  
Analysis Method ◽  
Key Factors ◽  
Heat Transfer Tube ◽  
Transfer Tube ◽  
Essential Components ◽  
Reliability Method

Abstract The heat transfer tube is one of the most essential components of the nuclear power plant as the boundary between the first and second circuit pressures. The wear between the heat transfer tube and the support plate or the anti-vibration strip is one of the essential reasons for its failure. Based on a heat transfer tube wear analysis method, combined with the reliability analysis theory, the calculation scheme of tube wear failure probability is proposed in this paper. In the analysis and calculation process, the key factors affecting the reliability are determined, including the baffle thickness B and the aperture difference Ce. In the manufacturing process, these key factors can be controlled, which is instructive for engineering practice.

Development of a Thermal Energy Storage System for Parabolic Trough Power Plants With Direct Steam Generation

2009 ◽  
Author(s):  
Doerte Laing ◽  
Thomas Bauer ◽  
Dorothea Lehmann ◽  
Carsten Bahl
Keyword(s):  
Heat Transfer ◽  
Power Plants ◽  
Storage System ◽  
Thermal Power ◽  
Sensible Heat ◽  
Steam Generation ◽  
Parabolic Trough ◽  
Two Phase ◽  
Solar Thermal Power ◽  
Direct Steam

For future parabolic trough plants direct steam generation in the absorber pipes is a promising option for reducing the costs of solar thermal power generation. These new solar thermal power plants require innovative storage concepts, where the two phase heat transfer fluid poses a major challenge. A three-part storage system is proposed where a phase change material (PCM) storage will be deployed for the two-phase evaporation, while concrete storage will be used for storing sensible heat, i.e. for preheating of water and superheating of steam. A pinch analysis helps to recognize interface constraints imposed by the solar field and the power block and describes a way to dimension the latent and sensible components. Laboratory test results of a PCM test module with approx. 140 kg NaNO3, applying the sandwich concept for enhancement of heat transfer, are presented, proving the expected capacity and power density. The concrete storage material for sensible heat was improved to allow the operation up to 500 °C for direct steam generation. A storage system with a total storage capacity of approx. 1 MWh is described, combining a PCM module and a concrete module, which will be tested in 2009 under real steam conditions around 100 bar.

Small Diameter Effects on Internal Flow Boiling

2001 ◽  
Author(s):  
Gail E. Kendall ◽  
Peter Griffith ◽  
Arthur E. Bergles ◽  
John H. Lienhard
Keyword(s):  
Heat Transfer ◽  
Thermal Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Flow Instability ◽  
Flow Boiling ◽  
Small Diameter ◽  
Internal Flow ◽  
Two Phase ◽  
Wide Range

Abstract Since the 1950’s, the research and industrial communities have developed a body of experimental data and set of analytical tools and correlations for two-phase flow and heat transfer in passages having hydraulic diameter greater than 6 mm or so. These tools include flow regime maps, pressure drop and heat transfer correlations, and critical heat flux limits, as well as strategies for robust thermal management of HVAC systems, electronics, and nuclear power plants. Designers of small systems with thermal management by phase change will need analogous tools to predict and optimize thermal behavior in the mesoscale and smaller sizes. Such systems include a wide range of devices for computation, measurement, and actuation in environments that range from office space to outer space and living systems. This paper examines important proceses that must be considered when channel diameters decrease, including flow distribution issues in single, parallel, and split flows; flow instability in parallel passages; manufacturing tolerances effects; nucleation processes; and wall conductance effects. The discussion focuses on engineering issues for the design of practical systems.

Dynamic Characteristics Analysis of Steam Generator Tube of Nuclear Power Plant by Finite Elements

2014 ◽  
Vol 577 ◽  
pp. 149-153
Author(s):  
Shuang Jiang ◽  
Jun Cai ◽  
Jing Wei Zhang ◽  
Qiao Zhi Sun ◽  
Xin Guo ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Modal Analysis ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Primary Circuit ◽  
Harmonic Load ◽  
Heat Transfer Tube ◽  
Finite Element Software ◽  
Transfer Tube

In nuclear power plants, the steam generator heat transfer tube is the weakest part of the primary circuit pressure boundary. Flow induced vibration is one of the main reasons for the failure of the heat transfer tube. In this paper, an ANSYS finite element software is used to carry out the modal analysis of the heat transfer tube, and to simulate the dynamic response of the heat transfer tube in the harmonic load based on the modal analysis.

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