Patent Number: 052251470
Section: summary

This application is accompanied by a microfiche appendix having 2 microfiche films. A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by any one of the patent disclosure, as it appears in the Patent and Trademark Office files or records, but otherwise reserves all copyright rights whatsoever. FIELD OF THE INVENTION This invention relates to analyzing light water reactor core neutronics in real-time, more specifically to determining core neutronics for simulation training and engineering analyzers. BACKGROUND OF THE INVENTION In the field of nuclear power facilities, it is important to analyze the reactor core neutronic properties for maintaining the nuclear power facility and training reactor operators to perform routine and emergency monitoring procedures. Heretofore, core neutronics for a light water reactor have been analyzed using engineering codes, such as the coarse mesh method described in Borresen, "A simplified, Coarse-Mesh, Three-Dimensional Diffusion Scheme for Calculating the Gross Power Distribution in a Boiling Water Reactor," Nucl. Sci. Engr., 44, 37, 1971, and methods of the RAMONA-3B code described in Wulff et al., "A Description and Assessment of RAMONA-3B Mod. O Cycle 4: A Computer Code with Three-Dimensional Neutron Kinetics for BWR System Transients," NUREG/CR-3664, Brookhaven National Laboratory, January 1984. These codes provide methods for providing a set of core neutronics parameters in a defined circumstance and analyzing or determining the resultant reactor core neutronics parameters in response to the given conditions. The Borresen reference refers to obtaining core neutronics data and solving modified two-group neutron diffusion equations for the two types of neutrons inside the core, namely the fast neutrons and the thermal neutrons. A thermal neutron may be considered as a fast neutron that has slowed down. More specifically, the reactor core is represented as a number of nodes that are spaced apart such that the fast neutrons have a relatively large mean free path (i.e., diffusion length) and the thermal neutrons have a low leakage from node to node. This permits using an approximation for the thermal neutrons leakage and a modification of the two-group equations to simplify the number of steps required to determine the core neutronics for the given conditions. RAMONA-3B, developed by the Brookhaven National Laboratory, uses the Borresen coarse mesh method and also relies on the fast neutrons as the determining criteria. However, the RAMONA-3B method relies on solving the two-group neutron diffusion equations by a finite difference method to determine the core neutronics for the given conditions. One of the problems with these known techniques is that the model does not have the capability to run the code from power plant start up to shutdown continuously in real time. They do not have the ability to analyze dynamic or static conditions in real-time. Consequently, they are limited in their application to selected transient conditions. Further, those known techniques are not sufficiently flexible to train operators under a wide variety of conditions or in real-time environments. It is therefore, an object of the present invention to provide for determining core neutronics in a real-time environment. It is another object to provide for a real-time analysis of core neutronics that can be used for simulation training of facility operators and for engineering analysis of core neutronics, separately or simultaneously. It is another object of the invention to provide for determining core neutronics in response to rapid transient conditions in a real-time environment. It is another object of the invention to provide a real-time analysis of core neutronics under normal and emergency operating conditions. It is another object of the invention to simulate real-time core neutronics under normal, emergency, and beyond design conditions continuously. SUMMARY OF THE INVENTION The present invention provides for methods and apparatus for sensing the core neutronic parameters of a nuclear reactor core and analyzing and determining the core neutronics in a real-time environment. One aspect of the present invention concerns real-time analysis of light water reactor core neutronics in detailed three-dimensional geometry. More specifically, a method is provided including modeling the reactor as a plurality of nodes in a conventional manner, monitoring the pertinent input core neutronics parameters for the type of reactor core, providing time dependent two group neutron diffusion equations that have been subjected to a space-time factorization of the neutron flux and delayed neutron precursors by amplitude and shape, substituting a coarse mesh finite difference approximation for fast neutron shape functions and determining the resulting core neutronics by application of the modified time dependent, two-group neutron diffusion equations, using a constant time step in the calculations. Preferably, the time step is not less than one quarter second. Another aspect of the invention concerns a method for determining the neutronics parameters of a reactor core. One such method comprises the steps of: representing the reactor core as a plurality of nodes; monitoring selected neutronic parameters of the reactor core; providing time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization by shape and amplitude functions in response to the plurality of nodes, sensing the monitored parameters; and determining the core neutronics parameters in response to the sensed parameters and the provided two group neutron diffusion equations in constant time steps, the time steps being less than one quarter second. Preferably, the method includes the step of selecting a coarse nodal representation of the reactor core and a time step for sensing the monitored parameters and determining the core neutronics parameters in a real-time environment. The solution methods in solving the shape functions and the amplitude functions are in real time, thereby providing the capability for simulating the full range of operation of a core continuously. The reference to the full range of operation of a core continuously should be understood to include, without limitation, transient, steady states, malfunction, and shutdown operations of the core. Preferably, for a pressurized water reactor, (PWR), each fuel assembly is represented as a radial node and for a boiling water reactor (BWR), each control cell having four fuel bundles surrounding a control blade position is represented as a radial node. Each radial node should have the same size. For BWRs, nodes on the core periphery will consist of fewer bundles. It should be understood that each radial node may have a plurality of radial nodes. Preferably the number of radial nodes is the same for each radial node, for example, from 8 to 24 axial nodes. This is known as a coarse node or coarse mesh model which permits use in a real-time environment. In a preferred embodiment, the invention also provides for monitoring certain thermohydraulic parameters associated with the reactor core, non-condensibles, and soluble boron quantities, and analyzing xenon and samarium concentrations and decay heat in the reactor core. The foregoing sensed thermohydraulic parameters are preferably provided by a real-time thermohydraulic analysis which is described in the copending and commonly assigned U.S. patent application Ser. No. 07/761,000, filed Sep. 17, 1991, entitled "REAL-TIME ANALYSIS OF POWER PLANT THERMOHYDRAULIC PHENOMENA", in the names of Guan-Hwa Wang and Zen-Yow Wang, which application is hereby incorporated by reference herein. This provides for simulator training and engineering analysis of a wide range of power plant scenarios, such as feedbacks between thermohydraulics and neutronics, operational and severe transients, human factor research, and design modification analysis. One advantage of the present invention is that it provides for analyzing a wide variety of fast transients, including, for example, thermohydraulic transients in the nuclear steam supply system, control rod movement, soluble boron changes, and xenon effects. Consequently, the invention can be used to simulate and to analyze core neutronics during startup and normal operation, anticipated operational occurrences, design-basis accidents, and many beyond design-basis accidents. Another advantage of the present invention is that it can be incorporated into a modern minicomputer or engineering workstation and utilized in a real-time environment. This provides for increased flexibility, particularly for simulation training of operators and real-time engineering analysis. Further, such a computer can be located in, near or remote from the control room of the reactor, thus providing for real-time simulation without interfering with the supervision or operation of the reactor. Another advantage of the invention is that it is compatible with many NRC-approved safety engineering analysis codes that are currently used for fuel management and reload safety analysis, thus providing for enhanced core neutronics analysis and simulation. The input data required by those engineering analysis codes can be easily adopted as the input data for the present invention.