Density-Wave Oscillations in Sodium Heated Once-Through Steam Generator Tubes

1981 ◽  
Vol 103 (3) ◽  
pp. 485-491 ◽  
Author(s):  
H. C. U¨nal
Keyword(s):  
Steam Generator ◽  
Empirical Relation ◽  
Density Wave ◽  
Horizontal Tube ◽  
Operating Conditions ◽  
Steam Quality ◽  
Water Steam ◽  
Percent Accuracy ◽  
Steam Generator Tubes ◽  
Density Wave Oscillations

Inception conditions of density-wave oscillations were determined in two sodium heated once-through steam generator tubes, i.e., in a 44.43 m long helical coil of 0.018 i.d. and in a tube of 0.0131 m i.d., comprised of a 9.40 m long vertical- and an 11.05 m long, V-shaped horizontal tube. The operating conditions on the water/steam side for the experiments were: pressure: 6–19.1 MN/m2; mass velocity 187–1020 kg/m2 s; inlet subcooling: 3.2–168 K; outlet steam quality: 1.15–2.38. The 306 data obtained and the 74 data found in the literature from sodium and electrically heated once-through steam generator tubes were correlated within 7.5 percent accuracy for 98 percent of the time. The R.M.S. error for all the 380 data is 3.33 percent. The observed density-wave oscillations are time-delay oscillations and the length of the superheated steam region and the transit time in this region practically govern the mechanism of these oscillations. An empirical relation has also been established for this mechanism.

Correlations for the Determination of the Inception Conditions of Density-Wave Oscillations for Forced and Natural Circulation Steam Generator Tubes

1980 ◽  
Vol 102 (1) ◽  
pp. 14-19 ◽  
Author(s):  
H. C. U¨nal
Keyword(s):  
Steam Generator ◽  
Mass Velocity ◽  
Natural Circulation ◽  
Density Wave ◽  
Steam Quality ◽  
Forced Circulation ◽  
Percent Accuracy ◽  
Steam Generator Tubes ◽  
Density Wave Oscillations ◽  
Heated Length

Accurate and simple correlations are presented to determine the inception conditions of density-wave oscillations in steam generator tubes. The correlations predict the power at the start of the density-wave oscillations within about 6.5 percent accuracy for long (i.e., L ≥ 10 m) forced circulation steam generator tubes and within about 20 percent accuracy for natural circulation and short forced circulation steam generator tubes. The ranges of the operating conditions and geometries for the data used to establish the correlations are as follows: Forced circulation tubes: Geometry: circular-straight tubes and serpentines, a circular coil and a rectangular straight tube; type of heating: electrical or sodium heating; the ratio of the heated length to diameter: 153–9502; pressure: 4.1–17.3 MN/m2; outlet steam quality: 0.27–1.85; inlet subcooling: 2.8–245.9 K; mass velocity: 118–2088 kg/m2s. Natural circulation tubes: Geometry and heating conditions: electrically heated circular tubes and annuli; ratio of the heated length to diameter: 34–489; pressure: 0.2–7.1 MN/m2; outlet steam quality: 0.04–0.62; inlet subcooling: 0–244 K; mass velocity: 529–1230 kg/m2s. The number of data considered is 106 for forced circulation tubes and 110 for natural circulation tubes.

Flow-Induced Vibration Model for Steam Generator Tubes in Two-Phase Flow

2008 ◽  
Author(s):  
Njuki W. Mureithi ◽  
Soroush Shahriary ◽  
Michel J. Pettigrew
Keyword(s):  
Force Field ◽  
Steam Generator ◽  
Two Phase Flow ◽  
Operating Conditions ◽  
Phase Flow ◽  
Two Phase Flows ◽  
Two Phase ◽  
Fluid Force ◽  
Vibration Stability ◽  
Steam Generator Tubes

While steam generators operate in two-phase flow, the complex nature of the flow makes the prediction of flow-induced fluidelastic instability of steam generator tubes a challenging problem yet to be solved. In the work reported here, the quasi-static fluid force-field, which is the important unknown for two-phase flows, is measured in a rotated-triangle tube bundle for a series of void fractions and flow velocities. The forces are shown to be strongly dependent on void fraction, flow rates and relative tube positions. The fluid force field is then employed along with quasi-steady vibration stability models, originally developed for single phase flows, to model the two-phase flow problem and predict the critical instability velocity. The results are compared with dynamic vibration stability tests and are shown to be in good agreement. The present work uncovers some of the complexities of the fluid force field in two-phase flows. The database provides new potential to designers to estimate expected fluid dynamic loads under operating conditions. The force field data may also be applied in dynamic computations for tube wear simulations, replacing the simple Connors’ model which is currently used.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Flow Stability of Heat Recovery Steam Generators

2004 ◽  
Vol 128 (4) ◽  
pp. 840-848 ◽  
Author(s):  
Heimo Walter ◽  
Wladimir Linzer
Keyword(s):  
Heat Recovery ◽  
Flow Instability ◽  
Natural Circulation ◽  
Density Wave ◽  
Horizontal Tube ◽  
Flow Stability ◽  
Steam Generators ◽  
Tube Bank ◽  
Density Wave Oscillations ◽  
Initial Results

This paper presents the results of theoretical flow stability analyses of two different types of natural circulation heat recovery steam generators (HRSG)—a two-drum steam generator—and a HRSG with a horizontal tube bank. The investigation shows the influence of the boiler geometry on the flow stability of the steam generators. For the two-drum boiler, the steady-state instability, namely, a reversed flow, is analyzed. Initial results of the investigation for the HRSG with a horizontal tube bank are also presented. In this case, the dynamic flow instability of density wave oscillations is analyzed.

PWR and WWER Steam Generator Tube Integrity Assessment

2013 ◽  
Author(s):  
Roman Krautschneider
Keyword(s):  
Steam Generator ◽  
Operating Conditions ◽  
Degradation Mechanisms ◽  
Steam Generator Tube ◽  
Integrity Assessment ◽  
Steam Generator Tubes

Paper is describing and comparing degradation mechanisms and integrity assessment of PWR and WWER type of steam generator tubes. Because of different design, different used materials and also different operating conditions, there are significant differences in degradation mechanisms. Therefore both steam generator types have their specific codes dealing with inspection, monitoring and maintenance.

Assessment of Dynamic Instability in Once Through Steam Generator

2008 ◽  
Author(s):  
V. Prakash ◽  
M. Thirumalai ◽  
P. Murugesan ◽  
V. Vinod ◽  
V. A. Sureshkumar ◽  
...  
Keyword(s):  
Water Flow ◽  
Steam Generator ◽  
Dynamic Instability ◽  
Flow Instability ◽  
Density Wave ◽  
Operating Conditions ◽  
System Control ◽  
Flow Oscillation ◽  
Steam Temperature ◽  
Stable Conditions

Hydrodynamic flow instability in Once Through Steam Generators (OTSG) is one of the important problems in the design and operation of Liquid Metal Fast Breeder Reactors (LMFBRs). Under certain operating conditions, water flow in OTSG is susceptible to instability due to the close coupling between the thermal and hydraulic processes. Sustained flow oscillations due to instability are undesirable since they result in flow mal-distribution among the tubes resulting in thermal stress, mechanical vibrations and system control problems. It is therefore, necessary to assess the operating conditions, under which instability occurs so that the system may be designed to operate always under stable conditions. The cause of the main type of instability, important for the design of SGs is the propagation of density waves. This type of low frequency instability is referred to in literature as parallel-channel, density wave, time delay or mass flow-void feedback instability. Dynamic instability (density wave oscillation DWO) occurs because of the phase mismatch between the primary perturbation (water flow) and the response to this perturbation (pressure drop). As many tubes are operating under essentially constant pressure heads, this mismatch can lead to sustained/diverging oscillations. Water flow oscillation in tubes manifests as oscillations in the steam temperature at the tube outlet/pressure fluctuations. However it is difficult to instrument individual tubes in SG for such measurement in an operating plant. If the flow oscillation in the tube manifests itself in the overall module flow, then fluctuation in the overall flow/flow noise could be utilized for on-line stability measurements. Towards this, experiments were conducted in the sodium heated once through steam generator in an OTSG model. To confirm the extent of oscillation in the steam temperature and in inlet water flow, 3 tubes out of 19, were monitored besides overall module flow. Main objective of the present study was to assess the occurrence of dynamic instability in SG through module inlet flow perturbations, measured by ΔP measurements across the orifice at entry to the tubes and steam temperature fluctuation measurement at the outlet of tube by bare thermocouples. This paper discusses the experiments carried out in the Steam Generator model of Prototype Fast Breeder Reactor to investigate the instability phenomenon, the instrumentation details, the results and its discussion.

Simulation Study of Dynamic Characteristic of Once-Through Steam Generator for CEFR

2016 ◽  
Author(s):  
Ye Shangshang ◽  
Yang Hongyi ◽  
Wang Xiaokun
Keyword(s):  
Dynamic Characteristic ◽  
Steam Generator ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Simulation Method ◽  
Nucleate Boiling ◽  
Simulation System ◽  
Stable Operation ◽  
Steam Quality ◽  
Water Steam

The reliability of steam generator is extremely important for the sodium-cooled fast reactor nuclear power plant safety and stable operation. The convective heat transfer mechanism of the once-through steam generator (OTSG) of China Experimental Fast Reactor (CEFR) was researched. The water/steam side was divided into four areas according to the heat flux and steam quality, named subcooled, nucleate-boiling, film-boiling, and superheater. In order to accurate determine the DNB, the CHF table was used in this paper. Based on the homogeneous flow model and fixed boundary method, a thermal-hydraulic simulation system, which named OTAC, was established in this paper. To evaluate its performance, the predictions of this method were compared with PSM-W code. The maximum difference between the temperatures predicted by this model and PSM-W was ∼5K. The calculated results are consistent with the actual experiment data, which indicates the correctness of the mathematical model and simulation method. Static and dynamic characteristic researches of CEFR OTSG have done in the simulation system. And the system can be used to simulate the OTSG dynamic in real-time.

scholarly journals Simulation System for PeLUIt 150 MW Nuclear Reactor by using vPower

2021 ◽  
Vol 2048 (1) ◽  
pp. 012034
Author(s):  
M Yunus ◽  
A A Budiman ◽  
S Zhe ◽  
Kiswanta ◽  
W Chunlin ◽  
...  
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Nuclear Power ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Operating Conditions ◽  
Transient Condition ◽  
Design Data ◽  
Water Steam ◽  
Main Steam

Abstract In developing the PeLUIt 150 MW nuclear power plant based on the High Temperature Gas-cooled Reactor (HTGR) technology, with the helium-coolant and output thermal power of 150 MW, the PeLUIt simulator is also developed for training the operators and educating other technical personnel. Referred to the balance of plant (BOP) design of the PeLUIt, the simulator utilized the vPower simulation platform to simulate the secondary loop for power generation with a water-steam Rankine cycle. The paper focuses on developing the secondary loop’s main components: steam generator, steam turbine, condenser, deaerator, and feedwater pump. The reactor module in the primary loop is simplified as a heat source with 150 MW output. The steam generator that connects the primary and secondary loops is modeled with the heat exchanger module by transferring heat from helium to water/steam. Meanwhile, pressure and flow parameters can also be simulated for both helium and water/steam flows in steady-state and transient operating conditions. The steady-state simulation results are almost the same as the design data. The differences in the main steam temperature, feedwater pressure, and feedwater temperature, are 0.03%, 0.53%, and 0.02%, respectively. Meanwhile, the transient condition carried out in the loss of coolant accident showed a decrease in flowrate of 43.31 kg/s and an increase in temperature of feedwater and main-steam of 52.32 and 15.38 °C, respectively. In addition, there was a pressure drop of around 10.37 (feedwater) and 10.16 MPa (main-steam).

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