套管式双面加热蒸汽发生器的热工水力分析

为研究套管式双面加热蒸汽发生器在稳态和瞬态过程中的热工水力特性，建立了描述蒸汽发生器物理现象的一维均匀流数学模型。应用该模型，开发了可计算稳态和瞬态工况下一回路和二回路冷却剂温度场、焓场的直流蒸汽发生器热工水力程序。计算结果对直流蒸汽发生器结构设计、运行具有指导意义。      

固定边界与移动边界直流蒸汽发生器模型的比较

介绍了一种直流蒸汽发生器的可移动边界模型。该模型按流体相的不同把蒸汽发生器划分为过冷区、两相区和过热区 3个部分。利用该模型编制了直流蒸汽发生器的瞬态热工水力分析程序。编制了固定边界模型的仿真程序。与试验数据及固定边界模型计算结果比较 ,表明可移动边界模型同时具有计算速度快和精度高的特点 ,是直流蒸汽发生器动态仿真的一种比较理想的模型     

蒸汽发生器优化设计后的运行特性研究

以秦山核电厂相关设备为原型,基于已开发的蒸汽发生器模型及优化计算程序,利用系统分析程序RELAP5验证该模型的准确性,并对优化设计所给出的蒸汽发生器的设计方案的稳态运行特性和负荷提升瞬态运行特性进行了模拟分析。结果显示：已开发的蒸汽发生器数学模型是合理的;在超负荷运行过程中,经优化设计的蒸汽发生器存在循环倍率过低问题;RELAP5可作为核动力设备优化设计方案的验证程序。     

采用漂移流模型的U形管蒸汽发生器动态仿真

使用漂移流模型建立U形管蒸汽发生器瞬态分析模型。在瞬态方程的基础上 ,得到稳态工况的计算方程 ,提出了一个计算速度快 ,精度高的稳态方程求解方法 ;其分析结果与大亚湾核电站蒸汽发生器静态工作特性符合得很好。在瞬态分析过程中 ,使用了自动变步长的非线性多步法 ,在保证计算精度的前提下 ,计算速度得以改善 ;其分析结果与经实验验证的FRAMATOME对大亚湾核电站蒸汽发生器分析结果符合得很好。     

整体式预热器U型管蒸汽发生器数学模型及动态模拟

本文针对整体式预热器Ｕ型管蒸汽发生器建立了热工数学模型并进行了动态模拟。该模型是一个２５阶、非线性、可移动边界模型。包括：一、二次侧的质量与能量平衡方程以及确定二次侧下降段流量的动量守恒方程。对所建立的模型利用Ｇｅａｒ方法作了蒸汽阀门开度阶跃１０％的瞬态模拟，所得结果和国外所公布的结果吻合得很好。由此可见。本文所建立的模型可以为这种蒸汽发生器的设计、运行、控制提供理论依据。     

基于两相漂移流理论的蒸汽发生器计算程序(CCM)的应用

介绍了一个建立在两相漂移流和相边界可连续移动模型基础上的、可较为全面描述单相及两相流的传热和流动的程序CCM.该程序通过"修正的有限元方法"对控制体节点之间建立联系,保证模拟效果的同时,节省了计算时间.以CCM为基础的UTSG-3程序可较好地模拟U型管蒸汽发生器的稳态和瞬态过程,可得到更为广泛的应用.通过不同运行参数的扰动研究了大亚湾核电站蒸汽发生器动态响应过程,结果显示UTSG-3程序具有模拟U型管蒸汽发生器动态行为的能力.     

钠冷快堆蒸汽发生器的模块化模型及瞬态仿真研究

研究了钠冷快堆电厂直管式直流蒸汽发生器可移动边界的模块化模型。采用ＭＡＴＬＡＢ语言,Ｃ语言编程与ＳＩＭＵＬＩＮＫ仿真工具相结合的建模方法,建立了仿真系统,对直管式直流蒸汽发生器的瞬态特性进行了研究,并获得了令人满意的结果。     

带螺旋缠绕管的双面加热管套管直流蒸汽发生器稳态换热研究

介绍了一种新型的内管为螺旋管的管套管式双面加热直流蒸汽发生器.为编写稳态换热分析程序建立了一种固定二次侧焓值边界计算模型.该模型依据二次侧流体相的变化,将蒸汽发生器划分为三个分区:单相液区、两相区和单相汽区.程序的计算结果与文献中实验结果吻合较好,并从计算结果分析得出了该蒸汽发生器的一些结构参数对换热效果的影响规律.     

核动力装置蒸汽发生器数值模拟计算

建立了一个合理完善的核动力装置蒸汽发生器动态特性分析数学模型，并运用Ｇｅａｒ方法对此模型求解。研制了用于核动力装置蒸汽发生器稳定运行及扰动和事故工况下动态数值模拟的安全分析程序。运用此程序对秦山核电厂蒸汽发生器失去给水的事故进行了计算，所得结果与大型程序ＲＥＬＡＰ－５计算结果符合较好，并对蒸汽发生器几种不同扰动序列的计算结果进行了理论分析。     

蒸汽发生器瞬态特性计算

1 前言 在核电站安全分析中,尤其在与二回路有关的事故分析中,蒸汽发生器瞬态特性计算十分重要。由于蒸汽发生器二次侧为两相流体,其运动和变化十分复杂,要精确地描述蒸汽发生器的瞬态特性,必需进行复杂的数学处理,这将给整个系统的瞬态分析带来困难。因此,在目前的系统安全分析中,作为一个部件的蒸汽发生器,一般都采用流体守恒方程的点模型来描述瞬态过程。     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Implementation of control rod movement and boron injection options by using control variables in RELAP5/PARCS V2.7 coupled code

In present neutron kinetics codes, control rods banks do not have the possibility of dynamic movement during the simulation of a transient; besides it is necessary to send the boron concentration from the thermal-hydraulic code to the neutronic code to account for changes in cross-sections due to boron dilution. For instance, control rod movements are pre-programmed with simple instructions introduced before the beginning of the calculation. Hence, control rod positions are not related to the core characteristics and the control systems at any time of the simulation. This work presents the changes introduced in RELAP5/PARCS v2.7 codes to achieve that control rods and the boron injection become more dynamic and realistic components in such kind of simulators. Furthermore, in order to test the modifications introduced in both codes, it has been analyzed a boron injection transient in a typical PWR Nuclear Power Plant. The thermal-hydraulic model includes all the primary loop components of a PWR, the core fuel assemblies modeled with PIPE components, pumps, steam generators, pressurizer, etc. The neutronic representation of the reactor has been made in a one-to-one basis fuel channel model for the whole core.     

Particle deposition and fouling in PWR steam generators

Commercial PWR steam generators have experienced reliability problems within the first decade of operation associated with material degradation, one of the causes of which is particle deposition and tube fouling. As a result steam generators often require costly outages for inspection and cleaning of fouling deposits. Knowledge of locations where sludge has accumulated in the steam generator can aid in planning and targeting locations for cleaning and removal of deposits. A particulate deposition model has been developed and implemented in the three dimensional thermal hydraulics computer code, ATHOS3 to calculate sludge and fouling regions within the steam generators during operation. This transient particle deposition model uses the thermal hydraulic field calculated by the ATHOS3 code, and the concentration of magnetite particles entering the steam generator to calculate the particle distributions and deposition on vertical and horizontal surfaces within the steam generator. Results of some simulations of operating steam generator designs are presented in this paper. These results show that preferred regions for deposition include hot side upper bundle and a kidney shaped region on top of the tube sheet.     

适用于微机的核蒸汽发生器热工水力分析程序—SGTH—2

本程序用于计算核蒸汽发生器的热工水力分布参数以及一次侧流动压降、二次侧自然循环和稳态特性,将本程序的计算结果与法国对同型号蒸汽发生器的实测数据以及用 ATHOS 程序的相应计算结果进行比较,表明主要热工水力参数能令人满意地吻合。     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code

This paper summarizes the current status of the Pennsylvania State University (PSU) version of the coupled three-dimensional (3-D) thermal-hydraulic/kinetics TRAC-PF1/NEM code for pressurized water reactor (PWR) transient and accident analysis and describes applications to reactivity insertion accident (RIA) simulations as well as recent developments. The TRAC-PF1/NEM methodology utilizes closely coupled 3-D thermal-hydraulics and 3-D core neutronics transient models to simulate the vessel and a 1-D simulation of the primary system. An efficient and flexible cross-section generation procedure was developed and implemented into TRAC-PF1/NEM. These features make the coupled code capable of modeling PWR reactivity transients, including boron dilution transients, in a reasonable amount of computer time. Three-dimensional studies on hot zero power (HZP) rod ejection and main steam line break (MSLB) transients in a PWR, as well as a large break loss-of-coolant-accident (LBLOCA) and boron dilution transients, were accomplished using TRAC-PF1/NEM. The results obtained demonstrate that this code is appropriate for analysis of the space-dependent neutronics and thermal-hydraulic coupled phenomena related to most current safety issues.     

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点,对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型,采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明,程序计算值与试验值趋势一致,最大相对偏差不超过4.34%,吻合程度较好。将验证后的程序与一回路系统程序耦合,分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况,得到了二回路系统的瞬态特性,为大型商用快堆电站的设计提供了参考。     

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is hours instead of the 5–6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less—about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of a one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the non-boiling and boiling region) and the necessary connecting coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), and perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding ‘open’ loop considerations, it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established.From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using the digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method.