Basınçlı su reaktörleri U borulu buhar üreteçlerinin termo-hidrolik modellenmesi
Thermal-hydraulic analysis of U-tube steam generators for gressurized water reactors
- Tez No: 16289
- Danışmanlar: PROF.DR. HASBİ YAVUZ
- Tez Türü: Doktora
- Konular: Nükleer Mühendislik, Nuclear Engineering
- Anahtar Kelimeler: Belirtilmemiş.
- Yıl: 1991
- Dil: Türkçe
- Üniversite: İstanbul Teknik Üniversitesi
- Enstitü: Fen Bilimleri Enstitüsü
- Ana Bilim Dalı: Belirtilmemiş.
- Bilim Dalı: Belirtilmemiş.
- Sayfa Sayısı: 271
Özet
ÖZET Basınçlı su reaktörleri'nde, nükleer güç çevrimini oluşturan en önemli dinamik sistemlerden birisi de buhar üreteçleri'dir. Nükleer güç çevrimi içersinde buhar üreteçleri, reaktör kalbi ile türbo- jeneratör gurubu arasında dinamik bağı oluşturmaktadır.. Santralın güvenirliliği ve emniyeti açısından buhar üreteçlerinin, kararlı ve geçici hallerdeki davranışlarının belirlenmesi gerekmektedir. Bu amaçla, sunulan çalışmada U-borulu buhar üreteçleri'nin dinamik davranışının detaylı termo-hidrolik analizi teorik olarak yapılmıştır. ü-borulu buhar üreteçleri, birincil kanalında reaktör kalbinden çıkan tek fazlı sıvı sıcak su ve ikincil kanalında bu sıcak sudan ısı alarak kaynayan iki fazlı akışkandan meydana gelmektedir. İkincil kanal, buhar tankı ile birlikte doğal dolaşım çevrimini oluşturmaktadır. Birincil kanalda akan tek fazlı akışkan tek fazlı akış modeline ve ikincil kanalda akan iki fazlı akışkan termal dengesiz drift-akiş modeline uygun olarak temel korunum ve bütünleyici denklemleri ile temsil edilmişlerdir. Bu çalışmada, temel korunum denklemleri, bindirmeli hücre modeli kullanılarak sonlu fark eşitliklerine, boru malzemesinde ısı iletimi ise sonlu elemanlar eşitliklerine dönüştürülmüştür. Tam kapalı formda yazılan sonlu fark eşitlikleri, geliştirilen üç basamaklı çözüm tekniği ve bilgisayar programları kullanılarak çözülmüştür. U-borulu buhar üreteçlerinde karşılaşılan en önemli kazalar; ana buhar borusu kırılması, birincil akışkan giriş sıcaklığının artması, türbin trip ve besi suyu kaybı kazalarıdır. Bu kazaların sonucunda, U-borulu buhar üreteçlerinde meydana gelen dinamik davranışların detaylı termo-hidrolik analizleri, geliştirilen bilgisayar programları yardımı ile yapılmış ve elde edilen teorik sonuçlar, nümerik ve grafikler halinde verilmiştir. Teorik sonuçlar ile, gerçek buhar üreteçlerinden alınan sonuçlar karşılaştırıldığında uyum içersinde oldukları saptanmıştır.
Özet (Çeviri)
THERMAL-HYDRAULIC ANALYSIS OF D-TOBE STEAM GENERATORS FOR PRESSURIZED WATER REACTORS SUMMARY The steam generators provide dynamic " links between the reactor core and turbine generator systems in pressurized-water-reactor (PWR) power plants. The ability to understand and predict the steady state and transient behavior of steam generators therefore is essential to the simulation and analysis of PWR plant response during load follow and anticipated transient conditions. Simulation and modeling of steam generator transients is then required both for safety assessment and for control system design. In case of severe anticipated transient conditions and accident analysis, such as turbine trip or main steam line break analysis, the requirements of steam generator models are significantly extended beyond those of operational plant transients. It is therefore necessary to perform detailed thermal and hydraulic analysis of the steam generators. Recent nuclear reactor accidents have indicated the necessity of understanding and prediction of the time dependent behavior of the nuclear reactor steam generators. There are few computer codes which have been developed so far specifically for nuclear steam generators. Indeed, the general purpose computer codes developed so far have proved not to be sufficient to give an accurate simulation to afore mentioned behavior. Therefore, the author was motivated to develop a flexible method to cover the overall behavior of the phenomena of ime dependent steam generation in nuclear reactor steam generators. PWR U-tube steam generators used in Westinghouse and Combustion Engineering provide heat sink for the primary nuclear heat source. The most prominent characteristic of the U-tube steam generator is the boiling two-phase flow, which involves phase separation in natural circulation and assymetric boiling. The understanding of the physical phenomena requires knowledge of flow regime, distribution, and natural circulation process. -IX-Thus, the purpose of this thesis are as follows: A. To understand physical description of the process taking place in steam generator unit. B. To develop a physical and a mathematical model using the basic concervation equations. C. To develop a numerical solution method which is stable. D. To develop a computer model for U-tube steam generator design. E. To compare the results of computer simulations with that of performance data obtained from nuclear plants and test facilities. U-tube steam generator (UTSG) is a vertical tube-and- shell heat exchanger with boiling secondary fluid and subcooled primary flow. In the tube bundle region, primary fluid (reactor coolant) flows inside the U-tubes upward first and then downward. On the secondary side, feedwater introduced into downcomer mixes with the recirculating water returning from the steam-water separator devices. In the heat exchanger region, heat is added to the water to produce a two phase mixture, which flow upward into the steam-water separator devices. The heat transfer regions for the secondary side of a UTSG unit are consist of single phase forced convection, nucleate boiling and the forced convection vaporization in normal operating conditions. The general fluid flow model developed in this study uses a finite difference formulation for basic conservation equations based on nonequilibrium drift-flux model. The basic field equations for nonequilibrium drift-flux model consists of two continuity equations and the mixture energy and momentum equations. In the state equation vapor phase is assumed to be at saturation. Thus, the state of water is expressed as functions of three independent properties, pressure, static quality and mixture enthalpy. The flow geometry is modeled using a one dimensional staggered spatial mesh description. The difference equations based on the concept of control volume result in defining mass and energy average properties at cell centers and requiring velocities at the cell boundaries. The spatial discretization is accomplished by integrating the fluid mass and energy equations over the cell volumes, and integrating the momentum equations across cell junctions.The spatial and time discretization of the fluid conservation equations is consistent with the basic extended implicit continuous-fluid Eulerian (EICE) method. The EICE method has been shown to provide stable numerical solutions in transient calculations for fluid flows ranging from nearly incompressible to highly compressible. This method is also capable of accounting for the fluid thermal expansion effects which occur in steam generator transients. In order to model noneguilibrium character of two phase flow, the author has developed nonequilibrium EICE method, abbreviated as the NEICE method. In the NEICE formulation, the fluid equation of state is represented by the first order expansion in density, pressure, static quality and enthalpy. The solution is obtained using a three-step iterative technique. The first step involves the coupled solution of the implicit mixture mass, momentum and state equations. In the second step, the coupled equations of mixture mass and energy are solved for the advanced-time entalpy distribution. The third step involves the solution of the coupled equations of mixture and vapor mass the for static quality. An essential feature which insures the stability of this solution method is the coupling of the fluid continuity equation with momentum, energy and vapor mass continuity equation. The NEICE method is applied in transient two-phase fluid flow calculations for secondary side of UTSG's. In the transient heat exchanger model the primary and secondary fluid flow solutions are calculated separately for each time step using previous-time values of heat flux at the tube surfaces. The advanced-time heat flux and temperature dependent tube metal properties are then calculated using the radial heat transfer model. The finite element formulation based on Galerkin method is used for the equations of metal heat conduction and metal to water heat transfer calculation. Empirical correlations for parallel flow to the rod bundles are mainly used in the calculation of heat transfer for single and two phase flows. The process of computation of the natural circulation in the UTSG is completed by coupling steam drum model with downcomer and riser regions. Two separate control volumes with a movable interface are used to represent vapor and liquid volumes in the steam drum. -XI-In this study, the models devoloped for single and two-phase flows and heat transfer are applied in UTSG designs. Using these models, a computer code is developed for steady state and transient simulations of steam generators. The steady state models are based upon solutions of the steady state forms Saf the finite difference equations used in transient fluid flow and heat conduction models. The steady state solutions are used to provide initial conditions for the transient calculations. Fluid boundary conditions required for the transient conditions include primary inlet temperature, pressure, flow rate, outlet pressure and steam drum feed inlet temperature, pressure and steam outlet flow rate. The steam outlet flow rate is also calculated by using a critical steam flow model which is available in the present computer code. The transient calculations of UTSG design correspond to a step increase in the primary temperature, in steam flow, and the turbine trip transient and complete loss of feed flow. The results of calculations in the above cases have been compared with the data obtained from plant performance or experimental test reports. Based upon these calculations the following main results and conclusions are made. 1. The numerical stability of the solution method is good for both mild and severe transients. 2. Steady state distributions of various parameters in primary and secondary sides are compared with the steady-state design parameters of UTSG. From this comparison, it is seen that steady- state calculations provides accurate prediction of initial conditions for transient calculations. 3. As a result of turbine trip, comparison of calculated values for steam drum pressure and feed chamber water level with experimental data shows a favorable prediction of the UTSG transient. It is also noted that computer calculations of drift-flux thermal equilibrium model are better agreement with experimental results. 4. The transient calculations in step increase in primary temperature show that the mechanism for the density wave oscillation involves coupling of water level, downcomer flow rate, riser flow rate and outlet quality. 5. In case of total loss of feedwater flow, the time for dryout of tubes are correctly predicted as compared with linear theory. -Xll-Based upon the results of this investigation it is concluded that models developed for UTSG in the present work can be applied for the study of UTSG performance during load follow and anticipated PWR plant transients. Furthermore, the NEICE fluid model developed in this study, can be applied for modeling other two-phase flow systems. Thus, models developed in this study can therefore be considered useful as a basis for further development of models for thermal and hydraulic analysis of fluid systems. -Xlll-
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