Steam Generator Blockage: A Thermal-Hydraulic Approach Based on CATHARE 2

2008 ◽  
Author(s):  
Alexis Vergnault ◽  
Robin Dorel ◽  
Fre´de´ric Goux
Keyword(s):  
Pressure Drop ◽  
Steam Generator ◽  
Tube Bundle ◽  
Power Level ◽  
Radiological Protection ◽  
Steam Generators ◽  
Flow Area ◽  
Flow Paths ◽  
Accident Scenarios ◽  
The Impact

In some French steam generators (SG) with broached tube support plates (TSP), the TSP flow paths have become blocked by deposits. Due to these blocked broached TSP, flow paths decrease the flow area and result in increased void fraction in the tube bundle region and pressure drop across the TSP. As a consequence, it also leads to a decrease of the recirculation ratio which can be observed through the SG level measurement. The major impacts of this phenomenon can be classified in three main categories: the mechanical impact on the TSP due to an increase of the pressure drop, the impact on the safety demonstration itself due to the less secondary side inventory that have to be considered for the thermal-hydraulic studies and finally the impact on the normal transients mainly due to the level oscillation encountered in the steam generator. This paper which focuses on the last item presents the Insitute for Radiological Protection and Nuclear Safety (IRSN) approach to calculate with CATHARE 2 [1] code the response of steam generators experiencing the blockage: both are considered transients with only a “stand alone” steam generator and transients performed with our SImulator for Post-Accident so called “SIPA” [2 to 4] which enables to simulate normal and accident scenarios. Sensitivity calculations on the blockage ratio, on the blockage distribution and on the power level are made to be able to assess this phenomenon. Some conclusions are arrived at the power levels that should be adopted to deal with blockage in order to restore the water level stability.

Proven Concepts for LLW-Treatment of Large Components for Free-Release and Recycling

2007 ◽  
Author(s):  
Lena Bergstro¨m ◽  
Maria Lindberg ◽  
Anders Lindstro¨m ◽  
Bo Wirendal ◽  
Joachim Lorenzen
Keyword(s):  
Steam Generator ◽  
Heat Exchangers ◽  
Tube Bundle ◽  
Pilot Project ◽  
Initial Weight ◽  
Steam Generators ◽  
Huge Number ◽  
Material Recycling ◽  
Waste Production ◽  
Sampling Program

This paper describes Studsvik’s technical concept of LLW-treatment of large, retired components from nuclear installations in operation or in decommissioning. Many turbines, heat exchangers and other LLW components have been treated in Studsvik during the last 20 years. This also includes development of techniques and tools, especially our latest experience gained under the pilot project for treatment of one full size PWR steam generator from Ringhals NPP, Sweden. The ambition of this pilot project was to minimize the waste volumes for disposal and to maximize the material recycling. Another objective, respecting ALARA, was the successful minimization of the dose exposure to the personnel. The treatment concept for large, retired components comprises the whole sequence of preparations from road and sea transports and the management of the metallic LLW by segmentation, decontamination and sorting using specially devised tools and shielded treatment cell, to the decision criteria for recycling of the metals, radiological analyses and conditioning of the residual waste into the final packages suitable for customer-related disposal. For e.g. turbine rotors with their huge number of blades the crucial moments are segmentation techniques, thus cold segmentation is a preferred method to keep focus on minimization of volumes for secondary waste. Also a variety of decontamination techniques using blasting cabinet or blasting tumbling machines keeps secondary waste production to a minimum. The technical challenge of the treatment of more complicated components like steam generators also begins with the segmentation. A first step is the separation of the steam dome in order to dock the rest of the steam generator to a specially built treatment cell. Thereafter, the decontamination of the tube bundle is performed using a remotely controlled manipulator. After decontamination is concluded the cutting of the tubes as well as of the shell is performed in the same cell with remotely controlled tools. Some of the sections of steam dome shell or turbine shafts can be cleared directly for unconditional reuse without melting after decontamination and sampling program. Experience shows that the amount of material possible for clearance for unconditional use is between 95 – 97% for conventional metallic scrap. For components like turbines, heat exchangers or steam generators the recycling ratio can vary to about 80–85% of the initial weight.

Steam Generator In-Service Inspections at Paks NPP

2008 ◽  
Author(s):  
Miklo´s Do´czi
Keyword(s):  
Heat Exchanger ◽  
Steam Generator ◽  
Eddy Current ◽  
Power Plants ◽  
Tube Bundle ◽  
Degradation Mechanism ◽  
Feed Water ◽  
Material Damage ◽  
Steam Generators ◽  
Service Inspection

Steam Generator is one of the most critical components in nuclear power plants. It has of overriding importance from point of view of safe and reliable operation of the whole plant. Variety of degradation mechanisms affecting SG tube bundle may cause different types of material damage. In Paks NPP eddy current in-service inspection have been performed since 1988. In the year 1997 higher number of defected tubes were found in case of Unit#2, compared to results of the previous years. A medium term SG inspection program had been performed in the time period between 1998–2004. Based on the results of eddy current inspections high number of heat exchanger tubes had been plugged. Chemical cleanings of all steam generators were performed aiming to reduce the magnetite, copper deposits and corrosion agents acting on the surface of the tubes. Replacement of the main condensers had been performed to stop the uncontrolled water income caused by the relatively frequent leakages of the condenser tubes. Several tube samples had been cut from the first row of the tube bundles of different steam generators to study the effectiveness of the cleaning process and to determine the composition of deposits on the tube outside surface. Also several tubes with eddy current indications had been pulled out from the steam generators to determine the acting degradation mechanism. Examination of removed tubes can provide opportunity to check the reliability of eddy current inspection using bobbin coil. Also there were tubes pulled out form SG with existing cracks. From the year 2005 new inspection program had been started. As the first results of the new inspection program shows, there is only a few new indications had been found and there is no measurable crack propagation in case of existing indications. During the recent years feed-water collectors were replaced in case of all units of the power plant, because of material damage (erosion corrosion). The paper summarizes the results of eddy current in-service inspection of heat exchanger tubes, results of examinations of removed tubes and also deals with results of visual examination of the feed-water distributor system.

Probabilistic Assessment of Fretting Wear in Steam Generator Tubes Under Flow Induced Vibrations

2003 ◽  
Author(s):  
Greg D. Morandin ◽  
Richard G. Sauve´
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Probabilistic Methods ◽  
Fretting Wear ◽  
Nonlinear Flow ◽  
Impact Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Damage ◽  
The Impact

Successful life management of steam generators requires an ongoing operational assessment plan to monitor and address all potential degradation mechanisms. A degradation mechanism of concern is tube fretting as a result of flow-induced vibration. Flow induced vibration predictive methods routinely used for design purposes are based on deterministic nonlinear structural analysis techniques. In previous work, the authors have proposed the application of probabilistic techniques to better understand and assess the risk associated with operating power generating stations that have aging re-circulating steam generators. Probabilistic methods are better suited to address the variability of the parameters in operating steam generators, e.g., flow regime, support clearances, manufacturing tolerances, tube to support interactions, and material properties. In this work, an application of a Monte Carlo simulation to predict the propensity for fretting wear in an operating re-circulation steam generator is described. Tube wear damage is evaluated under steady-state conditions using two wear damage correlation models based on the tube-to-support impact force time histories and work rates obtained from nonlinear flow induced vibration analyses. Review of the tube motion in the supports and the impact/sliding criterion shows that significant tube damage at the U-bend supports is a result of impact wear. The results of this work provide the upper bound predictions of wear damage in the steam generators. The EPRI wear correlations for sliding wear and impact wear indicate good agreement with the observed damage and, given the preponderance of wear sites subject to impact, should form the basis of future predictions.

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.

ARTIST: Introduction and First Results

2004 ◽  
Author(s):  
S. Gu¨ntay ◽  
D. Suckow ◽  
A. Dehbi ◽  
R. Kapulla
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Aerosol Deposition ◽  
Severe Accident ◽  
Steam Generators ◽  
Steam Content ◽  
Primary Particles ◽  
Accident Management ◽  
First Results ◽  
Small Flow

ARTIST (Aerosol Trapping In a Steam Generator) is a seven-phase international project (2003–2007) which investigates aerosol and droplet retention in a model steam generator under dry, wet and accident management conditions, respectively. The test section is comprised of a scaled steam generator tube bundle consisting of 270 tubes and 3 stages, one 1:1 separator unit, and one 1:1 dryer unit. As a prelude to the ARTIST project, four tests are conducted in the ARTIST bundle within the 5th EU FWP SGTR. These first tests address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities are sonic, and far away from the break, where the flow velocities are three orders of magnitude lower. With a dry bundle and the full flow representing the break stage conditions, there is strong evidence that the TiO2 aerosols used (AMMD 2–4 μm, 32 nm primary particles) disintegrate into much smaller particles because of the sonic conditions at the break, hence promoting particle escape from the secondary and lowering the overall DF, which is found to be between 2.5 and 3. With a dry bundle and a small flow reproducing the far-field velocities, the overall bundle DF is of the order of 5, implying a DF of about 1.9 per stage. Extrapolating the results of the dry tests, it turns out that for steam generators with 9 or more stages, it is expected that substantial DF’s could be achieved when the break is located near the tube sheet region. In addition, better decontamination is expected using more representative proxies of severe accident aerosols (sticky, multicomponent particles), a topic which is yet to be investigated. When the bundle is flooded, the DF is between 45 and 5740, depending on the mass flow rate, the steam content, and the water submergence. The presence of steam in the carrier gas and subsequent condensation inside the broken tube causes aerosol deposition and blockages near the break, leading to an increase in the primary pressure. This has implications for real plant conditions, as aerosol deposits inside the broken tube will cause more flow to be diverted to the intact tubes, with a corresponding reduction in the source term to the secondary.

Flow-Induced Wear in Steam Generator Tubes—Prediction Versus Operational Experience

1998 ◽  
Vol 120 (2) ◽  
pp. 138-143 ◽  
Author(s):  
M. K. Au-Yang
Keyword(s):  
Steam Generator ◽  
Tube Wall ◽  
Tube Bundle ◽  
Cross Flow ◽  
Operational Experience ◽  
Steam Generators ◽  
Support Plate ◽  
Power Operations ◽  
Full Power ◽  
Steam Generator Tubes

Many nuclear steam generators have accumulated more than 10 effective-full-power-years of operation. Eddy-current inspections revealed that a number of these steam generator tubes, notably those located in high local cross-flow regions, have indications of wear at some support plate elevations after 5 to 10 yr of effective-full-power operations. In the last 5 yr, a number of technical papers on nonlinear tube bundle dynamics has been published to address the effect of tube and support plate interactions. At the same time, test data relating wear and tube wall thickness losses for different material combinations and different support plate geometries became available. Based on the available data in the literature, as well as data obtained in the author’s affiliation, this paper assesses the cumulative tube wall wear after 5, 10, and 15 effective-full-power years of operation of a typical commercial nuclear steam generator, using different wear models. It is hoped that this study will shed some light on the probable mechanism that caused the observed wear in today’s operating nuclear steam generators.

Numerical Analysis on the Impact of Interstage Flow Addition in a High-Pressure Steam Turbine

2018 ◽  
Vol 140 (6) ◽  
Author(s):  
Soo Young Kang ◽  
Jeong Jin Lee ◽  
Tong Seop Kim ◽  
Seong Jin Park ◽  
Gi Won Hong
Keyword(s):  
High Pressure ◽  
Pressure Drop ◽  
Computational Analysis ◽  
Fluid Dynamic ◽  
Geometric Model ◽  
Flow Path ◽  
Complex Flow ◽  
Flow Paths ◽  
Single Passage ◽  
The Impact

This study analyzes the fluid dynamic characteristics of an ultrasupercritical (USC) high-pressure turbine with additional steam supplied through an overload valve between the second and third stages. The mixing between the main and admission flows causes complex flow phenomena such as swirl and changes of velocity vectors of the main flow. This causes a pressure drop between the second-stage outlet and third-stage inlet, which could potentially affect the performance of the turbine. First, a single-passage computational analysis, which is usually preferred in predicting the performance of multistage turbomachines, was performed using a simple model of an admission flow path and a single passage (SP) for the second and third stages of the turbine. However, the actual flow in the overload valve is supplied through the admission flow path, which has the shape of a casing that circumferentially surrounds the turbine, after flowing in two directions perpendicular to the turbine axis. This necessitates full-passage computational analyses of the two stages and the flow paths of the admission flow. To achieve this, we implemented a full three-dimensional (3D) geometric model of the admission flow path and conducted a full-passage computational analysis for all the flow paths, including those of the second and third stages of the turbine. The focus of analysis was on the pressure drop due to the admission flow. The results of the single and full-passage analyses were compared, and the effects of two different methods were analyzed.

Simulation of the Vibration of a Steam Generator Tube Subjected to Fluidelastic Forces Induced by Two-Phase Cross-Flow

2011 ◽  
Author(s):  
Teguewinde Sawadogo ◽  
Njuki Mureithi
Keyword(s):  
Steam Generator ◽  
Instantaneous Frequency ◽  
Tube Bundle ◽  
Cross Flow ◽  
Angle Of Incidence ◽  
Two Phase ◽  
Fluid Forces ◽  
Fluid Damping ◽  
Out Of Plane ◽  
The Impact

A computer simulation of the vibration of a tube bundle subjected to fluidelastic forces induced by two-phase cross flow is carried out. Two fluidelastic instability models are compared in the simulations: the Connors model and the quasi-steady model. In the quasi-steady model, the fluid forces are expressed in terms of the quasi-static drag and lift force coefficients and their derivatives which are determined experimentally. The consideration of the angle of incidence induced by the relative tube displacement with respect to the fluid introduces a velocity dependent term in the fluid force expression. The simulation has been done using ABAQUS. The ABAQUS user subroutine VUEL provides the required interface and information to calculate and apply the fluid forces to the structure. The fluidelastic forces applied to each element are calculated using the element displacement and velocity, the tube instantaneous frequency, a flow retardation parameter and the fluid damping and stiffness components. A typical U-tube in a steam generator subjected to a non-uniform two-phase flow was considered in the simulations. The tube support contact was modeled using the ABAQUS contact pair algorithm. The Anti-Vibration-Bars (AVB) limiting the tube vibration in the out-of-plane direction in the U-bend region were also included in this model. As the simulation was nonlinear because of the loose supports, a Fast Fourier Transform technique was used to estimate the tube instantaneous frequency. The quasi-steady fluidelastic force was compared to the fluidelastic force of the Connors model. The instability growth rate of the Connors model was found to be higher than that of the quasi-steady model. The impact forces at the supports and the AVB are also extracted. These forces can be used to calculate the work rate and estimate the tube wear.

Flow-Induced Vibration and Fretting-Wear Performance of CANDU™ Steam Generator U-Tubes: Instrumentation

2009 ◽  
Author(s):  
Atef Mohany ◽  
Victor Janzen
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Measurement Techniques ◽  
Fretting Wear ◽  
Process Conditions ◽  
Steam Quality ◽  
Steam Generators ◽  
Test Program ◽  
Two Phase ◽  
Support Design

This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU™ steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program is one of the initiatives that AECL is undertaking to demonstrate that the tube support design for future CANDU steam generators meets the stringent requirements associated with a 60 year lifetime. The main objective of the tests is to address the issue of in- and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with AVB supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper.

scholarly journals Performance Analysis and Tube Inlet Orifice Length Evaluation of a Once–Through Steam Generator

2019 ◽  
Vol 128 ◽  
pp. 03001
Author(s):  
Hun Sik Han ◽  
Han-Ok Kang ◽  
Juhyeon Yoon ◽  
Young In Kim ◽  
Youngmin Bae ◽  
...  
Keyword(s):  
Performance Analysis ◽  
Pressure Drop ◽  
Steam Generator ◽  
Flow Rate ◽  
Power Level ◽  
Thermal Power ◽  
Operating Conditions ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Tube Plugging

A numerical study is conducted for performance analysis and secondary side screw-type tube inlet orifice design of a once–through steam generator (OTSG). Various tube plugging conditions and power levels are considered, and the secondary coolant flow rate is adjusted to maintain a constant thermal power. Comprehensive numerical solutions are acquired to evaluate the OTSG thermal–hydraulic performance and minimum orifice length under various operating conditions. The OTSG performance is analyzed according to the tube plugging condition in terms of the OTSG thermal power, steam outletsuperheat degree, and secondary coolant pressure drop. The results obtained show that a constant thermal power canbe maintained by properly adjusting the secondary coolant flow rate with a variation ofthe steam outlet superheat degree and secondary coolant pressure drop when the OTSG operates at high power level. The required minimum orifice length to suppress the flow oscillation below the allowablelevelis evaluated. The lowest power level results in the highest minimum orifice length, and non-plugging condition provides a limiting case for the orifice length criterion.

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