scholarly journals Effect of helium purification system intake pipe on pressure drop of HTGR steam generator

2019 ◽  
Author(s):  
P. Kadarno ◽  
B. W. Riyandwita ◽  
Sriyono ◽  
I. D. Irianto
Keyword(s):  
Pressure Drop ◽  
Steam Generator ◽  
Intake Pipe ◽  
Purification System ◽  
System Intake

The Reasons for the Frequently Exceeded Pressure Drop of Resin Bed in Steam Generator Blowdown System

2017 ◽  
pp. 247-255
Author(s):  
Shunlong Yang ◽  
Heng Liu ◽  
Hui Liu ◽  
Huaqiang Wu ◽  
Xiangmeng Hu ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Steam Generator

scholarly journals Sensitivity Evaluation of AP1000 Nuclear Power Plant Best Estimation Model

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Drop ◽  
Temperature Variation ◽  
Steam Generator ◽  
Nuclear Power ◽  
Channel Number ◽  
Design Parameters ◽  
Coolant Temperature ◽  
Node Number

The best estimation process of AP1000 Nuclear Power Plant (NPP) requires proper selections of parameters and models so as to obtain the most accurate results compared with the actual design parameters. Therefore, it is necessary to identify and evaluate the influences of these parameters and modeling approaches quantitatively and qualitatively. Based on the best estimate thermal-hydraulic system code RELAP5/MOD3.2, sensitivity analysis has been performed on core partition methods, parameters, and model selections in AP1000 Nuclear Power Plant, like the core channel number, pressurizer node number, feedwater temperature, and so forth. The results show that core channel number, core channel node number, and the pressurizer node number have apparent influences on the coolant temperature variation and pressure drop through the reactor. The feedwater temperature is a sensitive factor to the Steam Generator (SG) outlet temperature and the Steam Generator outlet pressure. In addition, the cross-flow model nearly has no effects on the coolant temperature variation and pressure drop in the reactor, in both the steady state and the loss of power transient. Furthermore, some fittest parameters with which the most accurate results could be obtained have been put forward for the nuclear system simulation.

Large Eddy Simulation of a Helical Coil Steam Generator Test Section

2017 ◽  
Author(s):  
Jonathan K. Lai ◽  
Elia Merzari ◽  
Marilyn Delgado ◽  
Samuel J. Lee ◽  
Saya Lee ◽  
...  
Keyword(s):  
Pressure Drop ◽  
Large Eddy Simulation ◽  
Steam Generator ◽  
Test Section ◽  
Helical Coil ◽  
Reynolds Stresses ◽  
Eddy Simulation ◽  
High Heat Transfer ◽  
Large Eddy ◽  
Inlet Conditions

The helical coil steam generator (HCSG) is a compact heat exchanger that can have high heat transfer even when the pressure drop is low. This makes it advantageous in small modular reactors and high-temperature reactor designs. In order to investigate the fluid phenomena around these helical banked tubes, a test section was built at Texas A&M University to represent flow across two half-rods within HCSG. This study focuses on the validation of large eddy simulation (LES) for this particular geometry. Pressure tap and particle image velocimetry (PIV) measurements have been recorded at an inlet Reynolds number of 8643, and both mean and fluctuating data is compared with the numerical results. The highly scalable spectral-element code Nek5000 has been used to produce the LES calculations. First, simulations of varying polynomial order expansions are made to determine the spatial resolution required to capture the turbulent scales. Then, simulations with different inlet conditions are compared with experimental data. The pressure drop shows good agreement with pressure tap measurements while velocity shows similar characteristics with PIV. Furthermore, the components of the Reynolds stresses and modes from proper orthogonal decomposition have been developed to validate the physics captured.

Simple Nonlinear Methods for Predicting Two-Phase Instabilities in a Helically Coiled Steam Generator

2007 ◽  
Author(s):  
Seok-Ki Choi ◽  
Seong-O Kim ◽  
Han-Ok Kang
Keyword(s):  
Pressure Drop ◽  
Steam Generator ◽  
Flow Rate ◽  
Density Wave ◽  
Vertical Direction ◽  
Temporal Variations ◽  
Phase Region ◽  
Two Phase ◽  
Energy Agency ◽  
Proposed Model

A simple model to analyze the 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 the operating temperatures in the primary and secondary sides. The sizes of the three regions and 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.

Investigation on Hydraulic Resistance of Steam Generator Tube Support Plates

2014 ◽  
Author(s):  
Ting Xiong ◽  
Bo Wen ◽  
Yuanfeng Zan ◽  
Xiao Yan
Keyword(s):  
Experimental Data ◽  
Pressure Drop ◽  
Hydraulic Resistance ◽  
Steam Generator ◽  
Single Phase ◽  
Empirical Correlations ◽  
Two Phase ◽  
Flow Area ◽  
Pressure Drops ◽  
Phase Pressure

In order to obtain the hydraulic resistance characteristics of steam generator (SG) tube support plates (TSP), experimental as well as CFD studies have been carried out on both the single-phase and two-phase hydraulic resistances of various trefoil or quatrefoil orifice plates. Results show that with the increase of the Renylod number, the single-phase pressure drop coefficient decreases firstly and remains almost constant later. The single-phase pressure drop coefficient decreases with the increase of the chamfer radius of orifice or flow area. The two-phase pressure drops predicted by the empirical correlations are generally larger than the experimental results, while the pressure drops predicted by CFD software agree with the experimental data.

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