scholarly journals Fouling Of Nuclear Steam Generators: Fundamental Studies, Operating Experience and Remedial Measures Using Chemical Additives

2013 ◽  
Vol 2 (1) ◽  
pp. 61-88 ◽  
Author(s):  
C.W. Turner
Keyword(s):  
Economic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Operating Experience ◽  
Feed Water ◽  
Chemical Additives ◽  
Steam Generators ◽  
Primary Coolant ◽  
And Performance

Fouling remains a potentially serious issue that if left unchecked can lead to degradation of the safety and performance of nuclear steam generators (SGs). It has been demonstrated that the majority of the corrosion product transported with the feed water to the SGs accumulates in the SG on the tube-bundle. By increasing the risk of tube failure and acting as a barrier to heat transfer, deposit on the tube bundle has the potential to impair the ability of the SG to perform its two safety-critical roles: provision of a barrier to the release of radioactivity from the reactor coolant and removal of heat from the primary coolant during power operation and under certain post accident scenarios. Thus, it is imperative to develop improved ways to mitigate SG fouling for the long-term safe, reliable and economic performance of nuclear power plants (NPPs). This paper provides an overview of our current understanding of the mechanisms by which deposit accumulates on the secondary side of the SG, how this accumulation affects SG performance and how accumulation of deposit can be mitigated using chemical additives to the secondary heat-transport system. The paper concludes with some key questions that remain to be addressed to further advance our knowledge of deposit accumulation and how it can be controlled to maintain safe, economic performance of nuclear SGs.

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.

Development of High Performance Moisture Separator Reheater

2009 ◽  
Author(s):  
Issaku Fujita ◽  
Kotaro Machii ◽  
Teruaki Sakata
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
High Performance ◽  
Power Plants ◽  
Tube Bundle ◽  
Separation Performance ◽  
Heating Steam ◽  
Severe Erosion ◽  
Moisture Separator ◽  
And Performance

Moisture Separator Reheaters (MSRs) of Nuclear power plants, especially 1st generation type (commercial operation started from between 1970 and 1982), has been suffered from various problems like severe erosion, moisture separation performance deterioration, drain sub cooling. To solve these problems and performance improvement, improved MSR was developed. At the new MSR, high performance SS439 stainless steel round type tube bundle was applied, where heating steam distribution is optimized by orifice plate in order to minimize the drain sub cooling. Based on the CFD approach, cycle steam distribution was optimized and FAC resistant material application for the internal parts of MSRs was determined. As a result, pressure drop was reduced by 0.6% against the HP turbine exhaust pressure. Performance of moisture separation was improved by the latest chevron type separator. Where, the reverse pressure is locally caused at the drainage area of the separator because remarkable longitudinal pressure distribution is formed by the high-speed steam flow in the manifold. Then, a new moisture separation structure was developed in consideration of the influence that this reverse pressure gave to the separator performance.

Fluidelastic Instability Modeling of Loosely Supported Multispan U-Tubes in Nuclear Steam Generators

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Marwan Hassan ◽  
Atef Mohany
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Complex Dynamics ◽  
Tube Bundle ◽  
Design Guidelines ◽  
Fretting Wear ◽  
Contact Ratio ◽  
Steam Generators ◽  
Fluidelastic Instability ◽  
Wear Damage

Steam generators in nuclear power plants have experienced tube failures caused by flow-induced vibrations. Two excitation mechanisms are responsible for such failures; random turbulence excitation and fluidelastic instability. The random turbulence excitation mechanism results in long-term failures due to fretting-wear damage at the tube supports, while fluidelastic instability results in short-term failures due to excessive vibration of the tubes. Such failures may require shutdowns, which result in production losses, and pose potential threats to human safety and the environment. Therefore, it is imperative to predict the nonlinear tube response and the associated fretting-wear damage to tubes due to fluid excitation. In this paper, a numerical model is developed to predict the nonlinear dynamic response of a steam generator with multispan U-tubes and anti-vibration bar supports, and the associated fretting wear due to fluid excitation. Both the crossflow turbulence and fluidelastic instability forces are considered in this model. The finite element method is utilized to model the vibrations and impact dynamics. The tube bundle geometry is similar to the geometry used in CANDU steam generators. Eight sets of flat-bar supports are considered. Moreover, the effect of clearances between the tubes and their supports, and axial offset between the supports are investigated. The results are presented and comparisons are made for the parameters influencing the fretting-wear damage, such as contact ratio, impact forces, and normal work rate. It is clear that tubes in loose flat-bar supports have complex dynamics due to a combination of geometry, tube-to-support clearance, offset, and misalignment. However, the results of the numerical simulation along with the developed model provide new insight into the flow-induced vibration mechanism and fretting wear of multispan U-tubes that can be incorporated into future design guidelines for steam generators and large heat exchangers.

scholarly journals Analysis of Steam Generators Corrosion Products from Slovak NPP Bohunice

2012 ◽  
Vol 2012 ◽  
pp. 1-6 ◽  
Author(s):  
Jarmila Degmová ◽  
Július Dekan ◽  
Vladimír Slugeň ◽  
Constanze Thees ◽  
Ivan Smieško ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Construction Materials ◽  
Nuclear Industry ◽  
Feed Water ◽  
Secondary Circuit ◽  
Steam Generators ◽  
Transmission Geometry ◽  
Safety And Reliability ◽  
Main Components

One of the main goals of the nuclear industry is to increase the nuclear safety and reliability of nuclear power plants (NPPs). As the steam generator (SG) is the most corrosion sensitive component of NPPs, it is important to analyze the corrosion process and optimize its construction materials to avoid damages like corrosion cracking. For this purpose two different kinds of SGs and its feed water distributing systems from the NPP Jaslovske Bohunice were studied by nondestructive Mössbauer spectroscopy. The samples were scraped from the surface and analyzed in transmission geometry. Magnetite and hematite were found to be the main components in the corrosion layers of both SGs. Dependant of the material the SG consisted of, and the location in the system where the samples were taken, the ratios between magnetite and hematite and the paramagnetic components were different. The obtained results can be used to improve corrosion safety of the VVER-440 secondary circuit as well as to optimize its water chemistry regime.

Construction Aspects of the Doel 1 and 2 SGR Projects

2009 ◽  
Author(s):  
Bert Kroes ◽  
Edmond Gobert ◽  
Xavier Delhaye ◽  
Peter Devolder ◽  
Michel Sonville
Keyword(s):  
Nuclear Power ◽  
Steam Generators ◽  
Primary Coolant ◽  
Worker Exposure ◽  
Power Stations ◽  
Exchange Operation ◽  
Radiation Worker ◽  
Opening And Closing ◽  
Leak Before Break ◽  
Specific Construction

The Doel 1 and 2 PWR Nuclear Power Stations are the oldest commercially operating units in Belgium and the last to replace their two Steam Generators. The Doel 2 Steam Generators were replaced in 2004 and those of Doel 1 will be replaced late 2009. The replacement poses a particular challenge as these are the only stations in Belgium requiring the creation of primary and secondary containment opening for the SG exchange operation. Other construction challenges result from the a-typical SG support configuration which dates from the period well before the more or less standardized support configuration as used for later PWR units. The current paper discusses the construction approaches selected to facilitate the exchange operation and to minimize the outage duration and radiation worker exposure. The main particularities of the construction effort concern the secondary containment opening and closing using a structural formwork assembly, the use of containment platforms hanging inside the primary containment allowing for parallel primary and secondary containment reconstruction and the de-activation of some of the primary coolant piping and SG restraints following the licensing acceptance of the Leak Before Break concept for the primary piping. The specific construction options that made the Doel 2 replacement a success will be presented in this paper.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

Measurement of Flow-Induced Forces: A Single Flexible Tube in a Rigid Triangular-Pitch Bundle Subjected to Two-Phase Cross-Flow

2015 ◽  
Author(s):  
Enrico Deri ◽  
Joël Nibas ◽  
Olivier Ries ◽  
André Adobes
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Dynamic ◽  
Cross Flow ◽  
Maintenance Policy ◽  
Excitation Spectra ◽  
Two Phase ◽  
Force Coefficients ◽  
Out Of Plane

Flow-induced vibrations of Steam Generator tube bundles are a major concern for the operators of nuclear power plants. In order to predict damages due to such vibrations, EDF has developed the numerical tool GeViBus, which allows one to asses risk and thereafter to optimize the SG maintenance policy. The software is based on a semi analytical model of fluid-dynamic forces and dimensionless fluid force coefficients which need to be assessed by experiment. The database of dimensionless coefficients is updated in order to cover all existing tube bundle configurations. Within this framework, a new test rig was presented in a previous conference with the aim of assessing parallel triangular tube arrangement submitted to a two-phase cross-flow. This paper presents the result of the first phase of the associated experiments in terms of force coefficients and two-phase flow excitation spectra for both in-plane and out-of-plane vibration.

Digital I&C Operating Experience in the US

2008 ◽  
Author(s):  
Bruce Geddes ◽  
Ray Torok
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Failure Modes ◽  
Operating Experience ◽  
Data Set ◽  
Common Cause ◽  
Hardware Failure ◽  
Common Cause Failure ◽  
Defensive Measures

The Electric Power Research Institute (EPRI) is conducting research in cooperation with the Nuclear Energy Institute (NEI) regarding Operating Experience of digital Instrumentation and Control (I&C) systems in US nuclear power plants. The primary objective of this work is to extract insights from US nuclear power plant Operating Experience (OE) reports that can be applied to improve Diversity and Defense in Depth (D3) evaluations and methods for protecting nuclear plants against I&C related Common Cause Failures (CCF) that could disable safety functions and thereby degrade plant safety. Between 1987 and 2007, over 500 OE events involving digital equipment in US nuclear power plants were reported through various channels. OE reports for 324 of these events were found in databases maintained by the Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). A database was prepared for capturing the characteristics of each of the 324 events in terms of when, where, how, and why the event occurred, what steps were taken to correct the deficiency that caused the event, and what defensive measures could have been employed to prevent recurrence of these events. The database also captures the plant system type, its safety classification, and whether or not the event involved a common cause failure. This work has revealed the following results and insights: - 82 of the 324 “digital” events did not actually involve a digital failure. Of these 82 non-digital events, 34 might have been prevented by making full use of digital system fault tolerance features. - 242 of the 324 events did involve failures in digital systems. The leading contributors to the 242 digital failures were hardware failure modes. Software change appears as a corrective action twice as often as it appears as an event root cause. This suggests that software features are being added to avoid recurrence of hardware failures, and that adequately designed software is a strong defensive measure against hardware failure modes, preventing them from propagating into system failures and ultimately plant events. 54 of the 242 digital failures involved a Common Cause Failure (CCF). - 13 of the 54 CCF events affected safety (1E) systems, and only 2 of those were due to Inadequate Software Design. This finding suggests that software related CCFs on 1E systems are no more prevalent than other CCF mechanisms for which adherence to various regulations and standards is considered to provide adequate protection against CCF. This research provides an extensive data set that is being used to investigate many different questions related to failure modes, causes, corrective actions, and other event attributes that can be compared and contrasted to reveal useful insights. Specific considerations in this study included comparison of 1E vs. non-1E systems, active vs. potential CCFs, and possible defensive measures to prevent these events. This paper documents the dominant attributes of the evaluated events and the associated insights that can be used to improve methods for protecting against digital I&C related CCFs, applying a test of reasonable assurance.

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