Patent Number: 058870415
Section: summary

BACKGROUND OF THE INVENTION Field of the Invention This invention relates to a system for a nuclear power plant and, more particularly, to a system for identifying nuclear power plant components such as, for example, nuclear fuel assemblies. The invention also relates to a method for identifying nuclear power plant components. Background Information In a typical nuclear reactor, the reactor core includes a large number of elongated fuel assemblies. Conventional designs of these fuel assemblies include top and bottom nozzles, a plurality of elongated transversely spaced guide thimbles extending longitudinally between and connected at opposite ends to the nozzles, and a plurality of transverse support grids axially spaced along the guide thimbles. Each fuel assembly also includes a multiplicity of elongated fuel elements or rods. The fuel rods are transversely spaced apart from one another and from the guide thimbles. The transverse grids support the fuel rods between the top and bottom nozzles. The fuel rods each contain fissile material in the form of a plurality of generally cylindrical nuclear fuel pellets maintained in a row or stack thereof in the rod. The fuel rods are grouped together in an array which is organized so as to provide a neutron flux in the core sufficient to support a high rate of nuclear fission and, thus, the release of a large amount of energy in the form of heat. A typical nuclear reactor core contains about 100 to 200 nuclear fuel assemblies which are typically about 13 feet tall with a square cross-section having 8.5 inch sides. The fuel assemblies are vertically positioned in an array within the reactor core and are subject to both twisting and leaning motions away from their intended positions in the array. The top nozzle of each of the fuel assemblies has two locator holes on the top thereof. These locator holes must be properly aligned with corresponding locator pins of a reactor vessel head. The reactor vessel head, which normally rests on top of the reactor core, is typically made of 8 inch sheet steel and weighs about 10 to 15 tons. Therefore, it is critical that the fuel assemblies are suitably inline for correctly engaging the corresponding locator pins of the reactor vessel head before "dropping the head". During a refueling process, nuclear fuel assemblies are added to the reactor core and previously existing assemblies are either removed or repositioned. The correct positioning of the appropriate assembly is crucial in terms of reactor efficiency and safety. After the assemblies are placed in their planned positions, a video camera is swept over the entire core a plurality of times to verify the identity of each assembly and its correct placement relative to adjacent assemblies. The operator views the video output on a screen, reads a fuel assembly identifier or identification number engraved on a spring clamp of each assembly, and compares the identification number to a core map plan for verification. During this process, the operator dwells on each individual assembly until its identifier can be read. Light adjustments are made, if necessary, and unreadable identifiers are noted. In other camera sweeps, the camera is focused on the gaps between adjacent assemblies. The operator examines the video output and measures the distances between known points on adjacent assemblies to indirectly determine gap distances. At least three camera passes over the core are typical. The operator also compares the measured gaps to predetermined gap allowances for verification. It is known to manually use a measuring device, such as a ruler, on a video monitor in order to measure the distance or gap between nuclear fuel assemblies. However, such manual technique is laborious, subject to human error, and subject to cumulative errors as the various gaps are measured between all of the adjacent pairs of the fuel assemblies in the reactor core. Accordingly, there is room for improvement in determining the gap between an adjacent pair of fuel assemblies. It is also known to manually compare the identifier values and gap measurements with planned values for verification. Those values which cannot be determined with the video output are manually checked in the reactor core. Human mistakes may take much time to correct and require another verification before acceptance. Various difficulties are associated with the fuel assembly identity and gap alignment verification processes. Phenomena, such as poor lighting, the presence of shadows, and dark deposits, may make identifier values very difficult to read. Also, such phenomena may hide or distort key image features used for making gap measurements. In cases of uncertainty, the operator must physically go to the core and make a manual check, thereby being exposed to unwanted radiation. The current fuel assembly identity and gap verification processes together take approximately four to six hours to complete and is subject to human error. There is a need, therefore, for a method and apparatus which will significantly reduce verification time while increasing accuracy. SUMMARY OF THE INVENTION The invention is directed to an automated system for identifying at least one of a plurality of nuclear power plant components. The automated system includes camera means for inputting a first image of at least one of the nuclear power plant components and providing an input signal therefrom; digitization means for generating a second, digitized image of the nuclear power plant components from the input signal; means for locating the component identifier of the nuclear power plant components from pixel elements of the digitized image; and determining means at least for determining the component identifier including: plural recognizer means having an output providing an intermediate recognition of the component identifier, and means for combining the outputs of the plural recognizer means to recognize the component identifier. The outputs of the plural recognizer means may include an intermediate identifier and a corresponding confidence value. The means for combining the outputs may include means for determining unique ones of the intermediate identifiers, means for summing the confidence values corresponding to each unique intermediate identifier to provide a summed confidence value therefor, and means for recognizing the component identifier as the unique intermediate identifier having the largest summed confidence value.