Data processing: measuring – calibrating – or testing – Measurement system in a specific environment – Mechanical measurement system
Reexamination Certificate
2000-10-27
2002-11-12
Bui, Bryan (Department: 2863)
Data processing: measuring, calibrating, or testing
Measurement system in a specific environment
Mechanical measurement system
C702S050000, C702S100000, C073S861180
Reexamination Certificate
active
06480793
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
The present invention relates generally to flow condition monitoring systems and methods. In particular, the present invention relates to a system and method for monitoring flow conditions in the coolant and other fluids circulating through a nuclear power plant.
2. Description of the Related Art
In a nuclear plant of the pressurized water reactor (PWR) type, coolant fluid, which is basically boron and water, is continuously transferred through a closed circulation loop between a nuclear reactor and one or more steam generators.
During power production, the pressurized coolant absorbs heat released by the thermonuclear reaction occurring in the reactor. The heated coolant then flows through a main pipe which is appropriately known as the “hot leg” of the circulation loop. The hot leg delivers the hot coolant to a steam generator.
In the steam generator, the coolant fluid circulates through a heat exchanger. The heat exchanger cools the coolant fluid and uses the heat removed from the coolant to produce steam. This steam is eventually used to drive turbines and generate electricity.
After the circulating coolant is cooled by a heat exchanger, a circulation pump removes the coolant from the steam generator via a “suction leg” and returns it to the reactor via a “cold leg” and inlet. The coolant is then reheated in the reactor and the cycle repeats.
This circulation of coolant through one or more loops is critical for the operation of the power plant. Not only does it deliver heat energy to the steam generators where the energy is used to produce steam for driving the turbines, but the circulating coolant also prevents the reactor core in the reactor from overheating.
Nuclear power plant systems, including the steam generators, require periodic maintenance. In particular, the fluid circulation system must be inspected for potential degradation, and nozzle dams must be installed and removed from the steam generators to allow inspection and maintenance to be performed in a dry environment.
In order to install and remove nozzle dams, the coolant fluid must be drained from the steam generator. This requires lowering the fluid level in the main circulation loop and consequently the hot leg or main pipe. During such a maintenance period, which is termed a “shutdown,” the coolant continues to be heated by decay heat from the reactor core and is cooled by an alternate heat exchanger and auxiliary circulatory system known as the “shutdown cooling system.”
In order to lower the coolant or water level in the shutdown reactor system to permit refueling of the reactor core and to allow maintenance operations on portions of the system above the lowered water level, the water level must be controlled and maintained at a minimum level and flow rate to continuously provide adequate core cooling. This minimum level is about midway within the reactor coolant system main loop piping (the hot leg) and is commonly referred to as “midloop.”
During midloop operation, coolant water is circulated through the system to cool the core. Typically, there are a drain line or lines which communicate with the lower region of one or more of the main loop pipes or legs to draw the heated water from the core for cooling by the alternate heat exchanger in the shutdown cooling system and subsequent recirculation of cooled water to a reactor inlet and thus to the core.
It is possible to experience the formation of a Coriolis effect vortex in the drain line during midloop operation if the water level is lowered too far down or if the drain flow rate is too high. Such a vortex is undesirable because it limits the rate at which coolant flow can be drained from the system, and it can eventually lead to cavitations in the drain pump. Both results cause concern for continued cooling of the core.
The current methods to avoid vortex formation rely on keeping the water level as high as possible and/or reducing the flow rate, resulting in a conflict between the need to lower the water level for maintenance service, and the need to keep the water level high and at a sufficient rate for safe core cooling. Midloop measuring systems in use are related to a detection of the water elevation and inference of the status of the vortex therefrom.
In nuclear power plants much attention has been given to shutdown cooling system reliability, especially during reactor coolant system midloop water level operation. Midloop operation in a typical pressurized water reactor (PWR) nuclear steam supply system, for example, for the installation and removal of steam generator nozzle dams, can be a very difficult operational process. In fact, typically, the water level allowed tolerance is approximately plus or minus one inch (+/−1″). A vortex detection system has been disclosed in U.S. Pat. No. 5,861,560 by Robert P. Harvey to detect air vortexing and cavitation and thereby improve the shutdown cooling system reliability. However, the vortex detection system of Harvey is limited in its capability and usefulness because it relies only on the disruption of the signal of a conventional ultrasonic flowmeter to trigger an alarm indicating a vortex condition. The vortex detection system of Harvey is not capable or suitable for detecting various other fluid flow conditions throughout the nuclear reactor, such as fluid levels, entrained solid particulates caused by accident scenarios, condensible and noncondensible bubbles entrained in the fluid, and so forth. The vortex detection system of Harvey uses only one sensor and is looking only for the vortex condition in the drain pipe.
SUMMARY OF THE INVENTION
The present invention provides a flow condition monitor system and method for a nuclear reactor that rely upon acoustic detection of various flow conditions, including the existence of condensible or noncondensible bubbles entrained in the fluid, the existence and level of a free surface, the existence of vortex or whirlpool formations, the existence of entrained solid particulates, and various other flow conditions. The system uses a database of the acoustic characteristics of known flow conditions, and a processor that compares the detected acoustic signals with the known characteristics of the various flow conditions being monitored. The processor uses various means of discrimination, such as altering or decaying the transmitted signal, to aid in the interpretation, comparison and identification of the flow conditions.
The acoustic detection is provided by at least one sensor, and preferably a plurality of sensors, positioned to receive acoustic signals from the fluid flow being monitored. The sensor or sensors can be passive acoustic sensors, such as sensitive microphones or accelerometers attached to the pipe. Alternatively, the sensors can be ultrasonic devices that include ultrasonic transmitters and receivers positioned to capture variations associated with the coolant flow. In still another alternative, the sensors can be laser devices that include a laser source and a laser receiver diametrically opposed on a pipe structure whereby variations associated with the coolant flow cause unique disruptions in the laser signal.
In one embodiment, a first sensor is positioned upstream from a second sensor a sufficient distance that attenuations in the signal and noise detected by the first sensor can be detected by the second sensor. The signals can then be processed and compared with the acoustic characteristics of known flow conditions to determine the flow condition being detected. The transmitter and receiver of the sensors can be positioned on opposite sides of the pipe in which the coolant is flowing, or on the same side of the pipe, depending on the particular conditions and location of the fluid flow being monitored. An arrangement of the transmitter and receiver on opposite sides of the pipe will allow the compressibility difference of the water and air at the air/water interface to be taken into account, while an arrangement of the transmitter and receiver on the same side of
Bui Bryan
Westinghouse Electric Company LCL
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