Closed Model for Calculation of Differential Pressure After a Loss of Coolant Accident in Boiling Water Reactors

2009 ◽  
Author(s):  
Alexander Kratzsch ◽  
Wolfgang Ka¨stner ◽  
Rainer Hampel
Keyword(s):  
Differential Pressure ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Insulation Material ◽  
Closed Model ◽  
Loss Of Coolant Accident ◽  
Safety Research ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Residual Heat

The paper deals with the calculation of differential pressure on sieves after a loss of coolant accident (LOCA) in boiling water reactors. One of the main features in reactor safety research is the safe heat dissipation from the reactor core and the reactor containment of light-water reactors. In the case of loss of coolant accident the possibility of the entry of insulation material into the reactor containment and the building sump of the reactor containment and into the associated systems to the residual heat exhaust is a serious problem. This can lead to a handicap of the system functions. To ensure the residual heat exhaust it is necessary the emergency cooling systems to put in operation which transport the water from the sump to the condensation chamber and directly to the reactor pressure vessel. A high allocation of the sieves with fractionated insulation material, in the sump can lead to a blockage of the sieves, inadmissibly increase of differential pressure, build-up at the sieves and to malfunctioning pumps. Hence, the scaling and retention of fractionated insulation material in the building sump of the reactor containment must be estimated. This allows the potential plant status in case of incidents to be assessed. The differential pressure is the essential parameter for the assessment of allocation of the sieves.

Identification and Ranking of Phenomena Leading to Peak Cladding Temperatures in Boiling Water Reactors During Large Break Loss of Coolant Accident Transients

2002 ◽  
Author(s):  
Ruwan K. Ratnayake ◽  
S. Ergun ◽  
L. E. Hochreiter ◽  
A. J. Baratta
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Reactors ◽  
Computer Code ◽  
Boiling Water Reactors ◽  
Validation Process ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Accident Scenarios

In the licensing and validation process of best estimate codes for the analysis of nuclear reactors and postulated accident scenarios, the identification and quantification of the calculational uncertainty is required. One of the most important aspects in this process is the identification and recognition of the crucial contributing phenomena to the overall code uncertainty. The establishment of Phenomena Identification and Ranking Tables (PIRT) provides a vehicle to assist in assessing the capabilities of the computer code, and to guide the uncertainty analysis of the calculated results. The process used in this work to identify the phenomena was reviewing both licensing and best estimate calculations, as well as experiments, which had been performed for BWR LOCA analyses. The initial PIRT was developed by a group of analysts and was compared to existing BWR LOCA PIRTs as well as BWR LOCA analyses. The initial PIRT was then independently reviewed by a second panel of experts for the selected ranking of phenomena, identification of phenomena which were ignored, as well as the basis and rationale for the ranking of the phenomena. The differences between the two groups were then resolved. PIRTs have been developed for BWR types 4 and5/6 for the Large Break Loss of Coolant Accidents (LB-LOCA). The ranking and the corresponding rationale for each phenomenon is included in tables together with the assessed uncertainty of the code capability to predict the phenomena.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

The Steam Turbine Retrofitting of Six Boiling Water Reactor Units Using Innovative Engineering and Laser Scanning Techniques

2009 ◽  
Author(s):  
Mike Jones ◽  
David J. Nelmes
Keyword(s):  
Steam Turbine ◽  
Laser Scanning ◽  
Laser Scanner ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Extensive Experience ◽  
Turbine Generators ◽  
The Usa ◽  
Water Reactors ◽  
Scan Data

Alstom Power is executing the steam turbine retrofit of six nuclear units for Exelon Generation in the USA. The existing turbine-generators are an 1800 RPM General Electric design originally rated at 912 MWe and 1098 MWe and powered by Boiling Water Reactors. 18 Low Pressure inner modules will be replaced, with the first due to be installed in March 2010. This project is particularly challenging — the aggressive retrofit installation schedule is compounded by the requirement to handle radioactively contaminated equipment and also comply with demanding regulations applicable to BWR plant. The author’s company has extensive experience in the steam turbine retrofit business, having supplied around 800 retrofit cylinders globally since the 1970’s. However, this LP upgrade challenges the established techniques used in the business and requires extraordinary effort. Traditional retrofit engineering and installation principles have been interrogated and developed to meet the specific requirements of this project. Innovative techniques are introduced, including the extensive use of the Leica HDS 6000 laser scanner to model the existing plant. The approach has advanced the field of steam turbine retrofit design and installation significantly. The first section of this paper focuses on the extraordinary considerations of the project and the challenges surrounding BWR plant. The second part describes the laser scanning technique and the application of scan data. It outlines the innovative solutions which have been developed.

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