NUMEL-M: A COMPUTER CODE TO MODEL THE TRANSIENT RESPONSE OF A DRUM-TYPE STEAM GENERATOR

1982 ◽  
Author(s):  
D. Longworth ◽  
M.A. Walker ◽  
B. Scruton
Keyword(s):  
Steam Generator ◽  
Transient Response ◽  
Computer Code

A Nonlinear Dynamic Model of a Once-Through Subcritical Steam Generator

1976 ◽  
Vol 98 (3) ◽  
pp. 332-339 ◽  
Author(s):  
A. Ray ◽  
H. F. Bowman
Keyword(s):  
Power Plant ◽  
Steam Generator ◽  
Transient Response ◽  
Time Varying ◽  
Algebraic Equations ◽  
Process Variables ◽  
Nonlinear Dynamic Model ◽  
Phase Boundaries ◽  
Input Variables ◽  
Fundamental Equations

A dynamic thermal-hydraulic model of a once-through subcritical steam generator is presented which allows the investigation of power plant system transients. The three-section (economizer, evaporator, and superheater) model with time-varying phase boundaries is described by a set of nonlinear differential and algebraic equations derived from the fundamental equations of conservation of mass, momentum, and energy. The transient response of 8 process variables, due to 5 percent independent step disturbances in 5 input variables at 100 percent load, is discussed.

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

2018 ◽  
Author(s):  
Jun Huang ◽  
Junli Gou ◽  
Haifu Ma ◽  
Jie Fan ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Fluid Model ◽  
Computer Code ◽  
Medium Size ◽  
Hydraulic Characteristics ◽  
Pressurized Water

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

Development of a Computer Code, ONCESG, for the Thermal-Hydraulic Design of a Once-Through Steam Generator

2000 ◽  
Vol 37 (5) ◽  
pp. 445-454 ◽  
Author(s):  
Juhyeon YOON ◽  
Joo-Pyung KIM ◽  
Hwan-Yeol KIM ◽  
Doo Jeong LEE ◽  
Moon Hee CHANG
Keyword(s):  
Steam Generator ◽  
Computer Code ◽  
Hydraulic Design

Calculation of the Dose Load During Dismantling of Large Components in the Process of Decommissioning of Nuclear Installations

2013 ◽  
Author(s):  
Martin Hornacek ◽  
Vladimir Necas ◽  
Peter Bezak
Keyword(s):  
Steam Generator ◽  
Decay Time ◽  
Nuclear Power ◽  
Power Plants ◽  
Main Part ◽  
Computer Code ◽  
Point Of View ◽  
External Exposure ◽  
Dose Load ◽  
Type Reactor

The paper is focused on the decommissioning process from the point of view of steam generator dismantling. A brief description of the steam generator (used in nuclear power plants with VVER 440 type reactor) and used computer code VISIPLAN 3D ALARA are given. The main part deals with the created model and dismantling strategy together with variable input parameters — decay time and decontamination. The obtained results — external exposure of workers and the influence of time and pre-dismantling decontamination — are studied. Also detailed analyses of every dismantling step considered are presented.

Analysis of Two-Phase Flow Instability for Steam Generator

2008 ◽  
Author(s):  
Su-Xia Hou ◽  
Yun Tai ◽  
Fu-Yu Zhao
Keyword(s):  
Steam Generator ◽  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Flow Instability ◽  
Computer Code ◽  
Frequency Domain Method ◽  
Phase Flow ◽  
Steam Generators ◽  
Two Phase ◽  
The Stability

Two-phase flow instability is an important problem that affects the running of steam generators in nuclear reactor systems. In this paper, two-phase flow instability in parallel channels of a steam generator are analyzed to disclose the mechanism of flow instability by using the frequency domain method. The mathematical expressions of heat transfer and flow for a steam generator are proposed, and the transfer function of the closed-loop system is deduced by using linearization and Laplace transfer. The steam generator’s stability is judged according to Nyquist stability criterion. Depending on this fundamental principal, the computer code is developed to analyze the stability of steam generators. The results displayed two conclusions; firstly, the increase of inlet orifices or mass flow rate enhances the stability of generator steam; secondly, the coupling interactions between channels and their external loop effects on the stability of generator steam are not ignored. The result show that the effects are non-monotonic on the stability of generator steam.

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