Inspection of the Steam Generator Divider Plate

2010 ◽  
Author(s):  
Filippo D’Annucci ◽  
Eric Lecour
Keyword(s):  
Radiation Exposure ◽  
Steam Generator ◽  
Ultrasonic Inspection ◽  
Tube Sheet ◽  
Steam Generators ◽  
Inspection Techniques

The inspection of the steam generator divider plate becomes more and more important since indications have been detected in some steam generators in the area of the welds between tube sheet and divider plate. Remote operated manipulator and tools have been developed in order to minimize the radiation exposure of the personnel performing the inspections. Inspection techniques based on liquid penetrant, visual inspections have been developed and qualified to detect the indications. An ultrasonic inspection technique to determine the depth of the indications has been developed and qualified. The aim of this paper is to describe the inspection capabilities available to support the partition plate inspections.

Development of Deep Hole Drilling Technology for Tube Sheet of Steam Generator

2014 ◽  
Author(s):  
Shinichiro Ueda ◽  
Takuya Yagishita ◽  
Jun Maeno ◽  
Masanari Okuda
Keyword(s):  
Steam Generator ◽  
Control Technique ◽  
Hole Drilling ◽  
Drilling Machine ◽  
Tube Sheet ◽  
Deep Hole ◽  
Steam Generators ◽  
Deep Hole Drilling ◽  
Drilling Operation ◽  
Drilling Technology

Advanced PWR nuclear power plant has very-large steam generators of which tube-sheet is diameter 5000mm / thickness 500 mm / weight 100 ton or larger. Drilling of very thick tube sheet needs deep-hole drilling technique with strict dimensional tolerance control. In addition, the tube sheet has 10,000 or more tube holes, which means that stable drilling operation is mandatory for productivity control. Upon such background, steam generator tube sheet drilling operation needs precise dimensional control technique with highly efficient productivity. In this study, authors did (1) feasibility-study and development of deep hole drilling machine with the Boring and Trepanning Association (BTA) method, (2) cutting parameter evaluation such as feed rate and cutting velocity, (3) tube hole measurement system development, and (4) drill exchange program development with monitoring drilling machine motor loads and setting alarm level for the loads. Based on techniques developed by this study, IHI has achieved high-accuracy and stable deep-hole drilling technology for very-thick tube sheets applicable to the very-large steam generators.

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.

Ultrasonic Inspection of Liquid-Metal Fast Breeder Reactor Steam Generator Duplex Tubing

1980 ◽  
Vol 47 (2) ◽  
pp. 257-267
Author(s):  
K. J. Longua ◽  
G. K. Whitham ◽  
C. C. Allen
Keyword(s):  
Liquid Metal ◽  
Steam Generator ◽  
Ultrasonic Inspection ◽  
Fast Breeder Reactor

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.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

2002 ◽  
Author(s):  
April Smith ◽  
Kenneth J. Karwoski
Keyword(s):  
Steam Generator ◽  
Operating Experience ◽  
Alloy 600 ◽  
Steam Generators ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Water Reactors ◽  
1960S And 1970S ◽  
The 1960S ◽  
Thermally Treated

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.

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