Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

The use of high power density core is one of the promising ways to improve economic efficiency of advanced boiling water reactors. It is however known that in boiling two-phase flows, an increase in power density commonly reduces the margin to the onset of unanticipated flow instability. Hence, in the development of a boiling water reactor of high power density core, ability to predict the occurrence of boiling transition is considered indispensable even when the coolant flow rate is not in the steady state. In the present work, sinusoidal oscillation was applied to the inlet mass flux and the experimental measurement of the critical heat flux was carried out under flow oscillation conditions. It was shown that the critical heat flux decreases monotonically with increased values of oscillation amplitude and oscillation period. These results are consistent with experimental data reported by previous investigators. A simple theory was then proposed to estimate the critical heat flux in oscillatory flow condition. Considering the application to the advanced boiling water reactors, the triggering mechanism of the critical heat flux condition is supposed to be the liquid film dryout in annular two-phase flow regime of high vapor quality. Under the flow oscillation condition, it is expected that long waves are formed on a liquid film due to the time variation of inlet mass flux. Assuming that the wave evolution within a boiling channel is influential in the occurrence of the local dryout of a liquid film, an available nonlinear wave theory was applied to the estimation of critical heat flux under the flow oscillation condition. It was demonstrated that the critical heat fluxes measured under the oscillatory conditions agree with the proposed theory fairly well.

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Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

Volume 1: Plant Operations, Maintenance, Installations and Life Cycle; Component Reliability and Materials Issues; Advanced Applications of Nuclear Technology; Codes, Standards, Licensing and Regulatory Issues ◽

10.1115/icone16-48411 ◽

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.

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

ASME 2009 Power Conference ◽

10.1115/power2009-81140 ◽

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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