Patent Number: 
Section: description

The present invention may be implemented, for example, via FORTRAN computer program code executed using Digital Equipment Corp. Alpha computers running the Open VMS operating system. As embodied in a computer program for execution on a computer, the present invention determines actual power histories of each fuel rod in the reactor core using empirical data acquired from past operation of the reactor and evaluates the internal pressure for each fuel rod for an upcoming fuel cycle. Preferably, this rod evaluation process is performed during the fuel cycle design and licensing process for each operating cycle of a particular nuclear reactor. The evaluation process includes a rod-by-rod internal pressure analysis based on empirical data of actual operational power output levels of each fuel rod in the reactor core. A computer program constructs individual fuel rod power histories for each nuclear fuel rod in the reactor core based on information acquired during previous fuel cycles and a projected operation of the reactor in an upcoming fuel cycle. Using the constructed power histories for each fuel rod, the program then computes thermal and mechanical overpower limits and the maximum internal pressure for each rod in the upcoming fuel cycle. Licensing compliance is demonstrated by confirming that the computed maximum internal pressure for the upcoming fuel cycle is less than the critical pressure with a statistical confidence mandated by the regulatory agency and that the maximum thermal and mechanical overpower stresses of the fuel rods are below regulatory maximums. FIG. 1 shows a graph illustrating an example thermal-mechanical limit envelope 10 for evaluating fuel rods. The envelope represents the operating limit maximum linear heat generation rate (MLHGR) for a fuel rod as function of fuel pellet exposure. FIG. 2 shows a simplified block diagram of an example data processing system, 100, contemplated for performing the evaluation of fuel rod thermal-mechanical limits for each rod in a reactor core in accordance with the method of the present invention. Essentially, system 100 includes CPU 101, storage memory 102, and user interfacing I/O devices 103 and optionally one or more displays 104. Storage memory 102 includes a database or files (not shown) containing, for example, reactor plant initial state information, fuel lattice physics analysis results, 3D simulation results, fuel rod type/characteristics data and a program for evaluating fuel rods in accordance with the method of the present invention. FIG. 3 shows a mathematical relationship useful for computing an internal pressure ratio of a fuel rodxe2x80x94for producing pressure ratio values believed reliable to a 95% degree of confidence. The Pressure Ratio value obtained using this equation is based on the ratio between a maximum nominal internal pressure, Pmax,nom, and a nominal critical internal pressure, Pcrit,nom, for a fuel rod. Pmax,nom may be determined by performing a conventional T-M type analysis on a fuel rod, for example, as produced by GE""s GESTR performance software. FIG. 4 shows a functional program flow diagram of an example embodiment of the fuel rod evaluation program of the present invention. Each block of the diagram contains a concise explanation of a functional step performed, for example, by a computer program operating on a single or multi-processor computer system for the purpose of evaluating the thermal-mechanical limits for all fuel rods in a reactor core. One of ordinary skill will appreciate that the illustrated functional steps of FIG. 4, although explained in greater detail below, are essentially self explanatory and may be implemented on a conventional computer by utilizing conventional programming techniques and programming tools well known in the art. As illustrated by the functional flow diagram of FIG. 4, the method of the present invention essentially involves examining all fuel rods within each fuel bundle in a reactor core on a rod-by-rod basis to determine internal pressure data for each fuel rod and then using that information to set appropriate limiting operational criteria for the reactor. First, fuel rod power xe2x80x9chistoriesxe2x80x9d are constructed based on operating data from reactor process computers and data provided from pre-assembled data files (e.g., reactor 3D simulation files 304, reactor fuel lattice physics analysis files 306, and T-M analysis files 312). The reactor specific information provided by these data files may be pre-acquired and digitally stored by conventional means using standard procedures and processes well known in the nuclear industry. The fuel rod power histories are then used to perform T-M analysis for each rod individually. Referring to program flow diagram 300 of FIG. 4, fuel bundle xe2x80x9chistoriesxe2x80x9d, block 302, are first constructed using xe2x80x9chistoricalxe2x80x9d fuel-cycle data stored in a set of stored input files 304 (e.g., GE""s PANAEA CEDAR files). This historical fuel-cycle data is produced as a result of running 3D simulations of the reactor for operating conditions covering previous fuel cycles (e.g., cycle nxe2x88x923 through cycle nxe2x88x921) and the projected operating conditions for the upcoming cycle (cycle n). At this time, other reactor specific operational parameters relevant to constructing fuel bundle histories may be input as user data 301. The bundle histories so constructed may comprise, for example, fuel rod axial power rating (P), fuel rod exposure data (Exp), water density history (UH) and control fraction data (CF). Next, at block 308, individual fuel rod power histories are developed using the constructed fuel bundle history data and data obtained from a second set of input files 306 (containing reactor fuel lattice analysis data (e.g., GE""s TGBLA files). Next, at block 310, T-M analysis input files are constructed for running T-M analysis cases using a third set of input files 312 containing fuel rod specific data (e.g., GE""s GSTRM files). Next, at block 314, the T-M analysis cases for each rod are run and output files are produced. Finally, at block 316, the T-M output files are processed to provide output results, block 318, for each fuel rod for printing or display and then a next fuel bundle is examined. Output results 318 include at least peak pressure and exposure data for such rod and may further include other relevant information such as: end-of-life pressure/exposure, maximum exp./mode, maximum enrichment/GAD. This process continues until all bundles in the core are evaluated. FIG. 5 shows an example listing of rod-by-rod output results provided in the preferred implementation of the present invention.