• Nuclear Engineering and Technology
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• 제55권2호
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• pp.460-474
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• 2023

A simple but accurate mathematical model is crucial for dynamic simulations and controller design of helical coil once-through steam generator (OTSG). This paper presents a three-region movable boundary dynamic model of the helical coil OTSG. Based on the secondary side fluid conditions, the OTSG is divided into subcooled region (two control volumes), two-phase region (two control volumes) and superheated region (three control volumes) with movable boiling boundaries between each region. The nonlinear dynamic model is derived based on mass, energy and momentum conservation equations. And the linear model is obtained by using the transfer function and state space transformation, which is a 37-order model of five input and three output. Validations are made under full-power steady-state condition and four transient conditions. Results show good agreements among the nonlinear model, linear model and the RELAP5 model, with acceptable errors. This model can be applied to dynamic simulations and controller design of helical coil OTSG with constant primary-side flow rate.

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.343-352
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• 2006

Convenient analytical tools for evaluation of the aperiodic and the fluctuating instabilities of the passive residual heat removal system (PRHRS) of an integral reactor are developed and results are discussed from the viewpoint of the system design. First, a static model for the aperiodic instability using the system hydraulic loss relation and the downcomer feedwater heating equations is developed. The calculated hydraulic relation between the pressure drop and the feedwater flow rate shows that several static states can exist with various numbers of water-mode feedwater module pipes. It is shown that the most probable state can exist by basic physical reasoning, that there is no flow rate through the steam-mode feedwater module pipes. Second, a dynamic model for the fluctuating instability due to steam generation retardation in the steam generator and the dynamic interaction of two compressible volumes, that is, the steam volume of the main steam pipe lines and the gas volume of the compensating tank is formulated and the D-decomposition method is applied after linearization of the governing equations. The results show that the PRHRS becomes stabilized with a smaller volume compensating tank, a larger volume steam space and higher hydraulic resistance of the path $a_{ct}$. Increasing the operating steam pressure has a stabilizing effect. The analytical model and the results obtained from this study will be utilized for PRHRS performance improvement.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2395-2406
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• 2023

To study the influence of parameter uncertainty in small pressurized water reactor (SPWR) once-through steam generator (OTSG), the nonlinear mathematical model of the SPWR is firstly established. Including the reactor core model, the OTSG model and the pressurizer model. Secondly, a control strategy that both the reactor core coolant average temperature and the secondary-side outlet pressure of the OTSG are constant is adopted. Then, the uncertainty quantification method is established based on Latin hypercube sampling and statistical method. On this basis, the quantitative platform for parameter uncertainty of the OTSG is developed. Finally, taking the uncertainty in primary-side flowrate of the OTSG as an example, the platform application work is carried out under the variable load in SPWR and step disturbance of secondary-side flowrate of the OTSG. The results show that the maximum uncertainty in the critical output parameters is acceptable for SPWR.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.841-848
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• 2009

A simple model to analyze non-linear density-wave instability in a sodium-cooled helically coiled steam generator is developed. The model is formulated with three regions with moving boundaries. The homogeneous equilibrium flow model is used for the two-phase region and the shell-side energy conservation is also considered for the heat flux variation in each region. The proposed model is applied to the analysis of two-phase instability in a JAEA (Japan Atomic Energy Agency) 50MWt No.2 steam generator. The steady state results show that the proposed model accurately predicts the six cases of operating temperatures on the primary and secondary sides. The sizes of three regions, the secondary side pressure drop according to the flow rate, and the temperature variation in the vertical direction are also predicted well. The temporal variations of the inlet flow rate according to the throttling coefficient, the boiling and superheating boundaries and the pressure drop in the two-phase and superheating regions are obtained from the unsteady analysis.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.530-531
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.528-529
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• 2005

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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• pp.1664_1665
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• 2009

Many years ago, most of thermal power plants built in this country were of subcritical pressure, of medium or small size, of constant pressure operations and of drum type steam generators with circulation type boilers. But, nowadays almost all of them were of high efficiency, of supercritical pressure, of big capacity, of sliding pressure operations, and of once through type steam generator. It has such advantage as the reduction of startup duration, but it control system and operation method are very complicated. It has a big difference in operation method of turbine and boiler. The feedforward control needs to be introduced to prevent such problems as thermal shock during the transit from normal operation into bypass operation. This paper introduces the turbine control and turbine bypass control during startup of thermal power plants.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.398-403
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• 2001

Pressure drop at a helical square duct orifice is numerically evaluated. The orifice is installed at the entrance of a once-through steam generator tube to suppress flow instabilities. The calculated results are compared with the available experimental correlations, and showed good agreement. Effects of curvature ratio and characteristics of the secondary flow with Reynolds number are reported. Through the numerical simulations, pressure drop mechanisms were well understood inside the compact and complicated orifice geometry.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.623-628
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• 1998

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART Provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRDS of the SMART has excellent passive heat removal characteristics.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.