Thermal Stratification in Steam Generator Feedwater Lines

1984 ◽  
Vol 106 (1) ◽  
pp. 78-85 ◽  
Author(s):  
R. Braschel ◽  
M. Miksch ◽  
G. Schu¨cktanz
Keyword(s):  
Stress Intensity ◽  
Steam Generator ◽  
Thermal Stratification ◽  
Load Case ◽  
Steam Generators ◽  
Design Modification ◽  
Material Fatigue

In recent years cracks have been found in the inlet nozzles of feedwater lines of steam generators. These cracks occur as a result of material fatigue. Thermal stratification in the feedwater is probably one of the factors primarily responsible for the fatigue. This paper describes methods of calculating the stress intensity ranges occurring as a result of the stratification. In addition, a design modification (a diversion tank) is proposed which effectively prevents the occurrence of this load case.

Stress and Fatigue Analysis Considering Thermal Stratification With the THERMAXS Code

2009 ◽  
Author(s):  
Adama Diaby ◽  
Michel De Smet
Keyword(s):  
Stress Intensity ◽  
Steam Generator ◽  
Thermal Stratification ◽  
Water System ◽  
Computer Code ◽  
Fatigue Analysis ◽  
Feed Water ◽  
Steam Generators ◽  
Thermal Ratcheting ◽  
Division 1

This paper describes the stress and fatigue analysis of the feed water nozzles of the replacement steam generators of the Doel 2 NPP in Belgium. In the framework of the steam generators replacement, thermal stratification transients were considered in the stress and fatigue analyses of the feed water system components such as the feed water lines, their reactor building penetrations and the steam generator feed water nozzles. To do so, long term external wall temperature measurement have been performed on one feed water line of Doel 2 between the steam generator replacement in 2004 and December 2007. From those measurements a number of typical stratification phenomena were identified. For each phenomenon, a design stratification transient and its number of occurrences were derived from the measurements. A fatigue analysis and primary+secondary stress intensity analysis of the feed water nozzles of the Doel 2 NPP was performed taking into account the presence of the design thermal stratification transients in the feed water lines and feed water nozzles. The fatigue analysis was performed according to the rules of the ASME Boiler & Pressure Vessel Code, Section III, Division 1 – Subsection NB-3200 using the in-house developed computer code THERMAXS that is capable of dealing with measured stratification transients. For the analysis, one may finally conclude that the fatigue, P+Q stress intensity range and thermal ratcheting criteria are respected throughout the 40 years of plant life.

On the Assessment of Through-Wall Circumferential Cracks in Steam Generator Tubes With Tube Supports

2003 ◽  
Vol 125 (1) ◽  
pp. 85-90 ◽  
Author(s):  
Xin Wang ◽  
Wolf Reinhardt
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Steam Generator ◽  
Limit Load ◽  
Life Extension ◽  
Steam Generators ◽  
Failure Assessment ◽  
Circumferential Defects ◽  
Steam Generator Tubes

The assessment of steam generator tubes with defects is of great importance for the life extension of steam generators. Circumferential through-wall cracks are the most severe of all tube circumferential defects, and usually require plugging of the affected tubes. The assessment of the tubes with through-wall circumferential cracks or cracks projected to become through-wall can be conducted using the failure assessment diagram (FAD) approach. This approach requires the calculation of the stress intensity factor and the limit load. The available stress intensity factor and limit load solutions for cracked tubes do not include the constraining effect of the tube supports. In the present paper, it is shown that this can be overly conservative. Solutions for stress intensity factors and limit loads are presented for tubes with circumferential through-wall cracks including the effect from the tube support plates. Different values of support spacing are considered. Based on these solutions, the assessment of a typical steam generator tube is demonstrated.

Thermal Fatigue Analysis of a PWR NPP Steam Generator Injection Nozzle Model Subjected to the Thermal Stratification Phenomenon

2009 ◽  
Author(s):  
Luiz Leite da Silva ◽  
Tanius Rodrigues Mansur ◽  
Carlos Alberto Cimini
Keyword(s):  
Fatigue Limit ◽  
Steam Generator ◽  
Thermal Fatigue ◽  
Thermal Stratification ◽  
Stratified Flows ◽  
Experimental Results ◽  
Pipe Material ◽  
Injection Nozzle ◽  
Experimental Section ◽  
Material Fatigue

This work is related to an experimental thermal stratification study aiming to quantify thermal fatigue damages in the pipe material. Thermal fatigue damages appear as a consequence of non-linear longitudinal and circumferential loads and thermal stripping present in pipes whit thermal stratified flows. In this work an experimental section, simulating the injection nozzle of a NPP steam generator, was subjected to the effects of thermal fatigue due to thermal stratification. The experimental section was made of stainless steel pipe type AISI 304L and its geometric characteristics allowed the same range of Froude numbers of a Pressurized Water Reactor (PWR) NPP. Temperatures were measured externally and internally in three positions and deformations just externally in seven positions. Up-and-down fatigue tests were done to assess the amount of damage induced in the material experimental section. Preliminary numerical simulations were done using a coupled analysis in the ANSYS code with temperatures and pressure inputs taken from thermo hydraulic experimental results. The objectives in this work are quantify the thermal fatigue intensity imposed to the pipe material by thermal stratification experiments, verify the agreement between numerical and experimental thermal stratification results and obtain the material fatigue limit testing specimens made of pipe experimental section and from the virgin pipe. In this work is possible to conclude that stratified flows could be developed in the experimental section, thermal stratification induces considerable thermal stresses and strains in the experimental section pipe material, thermal stratification reduces the material fatigue limit, numerical and experimental results agreed appropriately in some pipe region and disagreed in others.

Parametric Study of External Pressure on Steam Generator Tube Bends

2012 ◽  
Author(s):  
Mitch Hokazono ◽  
Clayton T. Smith
Keyword(s):  
Steam Generator ◽  
Parametric Study ◽  
External Pressure ◽  
Light Water Reactor ◽  
Material Strength ◽  
Steam Generators ◽  
Elliptical Cross ◽  
Light Water ◽  
Water Reactor ◽  
Steam Generator Tubes

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

scholarly journals Gas Turbine Heat Recovery Steam Generators for Combined Cycles Natural or Forced Circulation Considerations

1988 ◽  
Author(s):  
Akber Pasha
Keyword(s):  
Gas Turbine ◽  
Steam Generator ◽  
Heat Recovery ◽  
Power Plants ◽  
Natural Circulation ◽  
Combined Cycle ◽  
Heat Recovery Steam Generator ◽  
Steam Generators ◽  
Forced Circulation ◽  
Gas Turbine Exhaust

In recent years the combined cycle has become a very attractive power plant arrangement because of its high cycle efficiency, short order-to-on-line time and flexibility in the sizing when compared to conventional steam power plants. However, optimization of the cycle and selection of combined cycle equipment has become more complex because the three major components, Gas Turbine, Heat Recovery Steam Generator and Steam Turbine, are often designed and built by different manufacturers. Heat Recovery Steam Generators are classified into two major categories — 1) Natural Circulation and 2) Forced Circulation. Both circulation designs have certain advantages, disadvantages and limitations. This paper analyzes various factors including; availability, start-up, gas turbine exhaust conditions, reliability, space requirements, etc., which are affected by the type of circulation and which in turn affect the design, price and performance of the Heat Recovery Steam Generator. Modern trends around the world are discussed and conclusions are drawn as to the best type of circulation for a Heat Recovery Steam Generator for combined cycle application.

Crack Growth Prediction and Leakage Potential of Steam Generator Tubes Subjected to Flow Induced Vibrations

2014 ◽  
Author(s):  
Salim El Bouzidi ◽  
Marwan Hassan ◽  
Jovica Riznic
Keyword(s):  
Crack Growth ◽  
Steam Generator ◽  
Power Plants ◽  
Mechanical Energy ◽  
Steam Generators ◽  
Two Phase ◽  
Element Analysis ◽  
Wide Range ◽  
Flow Induced Vibrations ◽  
Crack Growth Model

Nuclear steam generators are critical components of nuclear power plants. Flow-Induced Vibrations (FIV) are a major threat to the operation of nuclear steam generators. The two main manifestations of FIV in heat exchangers are turbulence and fluidelastic instability, which would add mechanical energy to the system resulting in great levels of vibrations. The consequences on the operation of steam generators are premature wear of the tubes, as well as development of cracks that may leak radioactive heavy water. This paper investigates the effect of tube support clearance on crack propagation. A crack growth model is used to simulate the growth of Surface Flaws and Through-Wall Cracks of various initial sizes due to a wide range of support clearances. Leakage rates are predicted using a two-phase flow leakage model. Non-linear finite element analysis is used to simulate a full U-bend subjected to fluidelastic and turbulence forces. Monte Carlo Simulations are then used to conduct a probabilistic assessment of steam generator life due to crack development.

Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig

2008 ◽  
Author(s):  
A. D. Efanov ◽  
S. G. Kalyakin ◽  
A. V. Morozov ◽  
O. V. Remizov ◽  
A. A. Tsyganok ◽  
...  
Keyword(s):  
Steam Generator ◽  
Large Scale ◽  
Test Procedure ◽  
Heat Removal ◽  
Primary Circuit ◽  
Test Rig ◽  
Steam Generators ◽  
Main Equipment ◽  
Mode Of Operation ◽  
Condensation Mode

In new Russian NPP with VVER-1200 reactor (V-392M reactor plant) in the event of an accident being due to the rupture of the reactor primary circuit and accompanied by the loss of a.c. sources, provision is made for the use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. To experimental investigation of the condensation mode of operation of VVER steam generator, a large scale HA2M-SG test rig was constructed. The test rig incorporates: tank-accumulator, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger simulator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure – 1.6 MPa, temperature – 200 Celsius degrees. Experiments at the HA2M-SG test rig have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond basis accident. The report presents the test procedure and the basic obtained test results.

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.

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