A method and system for demonstrating compliance of nuclear fuel rods with fundamental licensing criteria for fuel rod internal pressure during nuclear reactor operation is presented. A nuclear fuel 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 empirical information acquired during previous fuel cycles and the 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 a maximum internal pressure for each rod in the upcoming fuel cycle. Licensing compliance is demonstrated by confirming that the computed maximum internal pressure and maximum thermal and mechanical overpower stresses of any fuel rod for the upcoming fuel cycle is less than the fundamental licensing limits.

A nuclear reactor core performance data visualization system provides a method and apparatus for extracting and visually displaying large amounts of numerical performance data acquired from an operational nuclear reactor or from a nuclear reactor core simulator. The visualization system apparatus includes a color display device, a digital computer having a data storage memory, and one or more data communications links for communicating with reactor core sensors and/or other computers. After acquiring performance parameter data, the visualization system computer extracts selected performance data and constructs two distinct data arrays in memory that are used in creating formatted files for generating specific types of visual displays. A first data array, containing time-varying reactor performance parameter data corresponding to the performance of reactor core elements in two dimensions over a predetermined duration of time, is used, to create digital image animation files for displaying dynamic color-coded graphics, including reactor core element diagrams and multi-ordinate graphs. Once animated, the displayed graphics allow the user to view a time-wise evolution of the changing numerical values for selected core parameters and other time-dependent variables. A second data array, containing spatially-related core performance parameter data corresponding to performance of reactor core elements in three dimensions at one or more selected point(s) in time, is used to create virtual reality modeling language (VRML) files for displaying a virtual 3-D color-coded reactor core model that can be manipulated in three dimensions by the viewer.

The exit thermocouples of a pressurized water reactor (PWR) are calibrated by recording the thermocouple temperature measurements periodically during power ascension of the reactor together with a predicted power at the corresponding locations at the same time determined by a three-dimensional analytical nodal model of the core at the same core average power. The temperature measurements are converted to local core power values which are then divided into the corresponding predicted powers for the corresponding locations to arrive at mixing factors which are fitted to a selected function of core power. These mixing factors are recorded and subsequently applied to local power values calculated from measured exit thermocouple temperatures to adjust the three-dimensional nodal model power. Periodically, the mixing factors are adjusted by using flux map data to update the three-dimensional analytical nodal model power to generate a reference power distribution, calculating updated mixing factors using the current thermocouple temperature measurement taken at the same time and therefore the same average core power, and the reference power distribution, and then shifting the mixing factor functions of core power accordingly to pass through the updated mixing factors.

An incore monitoring method of a nuclear reactor, includes, measuring neutron flux levels at pitch levels corresponding to local power range monitor sensors arranged along an axial direction inside a detector assembly installed in a nuclear reactor; performing power calculation, including calculation of thermal characteristics, of fuel assembly group consisting of fuel assemblies adjacent to the corresponding detector assembly, based on indicated values of the local power range monitor sensors of the corresponding detector assembly at a first time, calculating thermal characteristics at a second, subsequent time in which the power calculation is not calculated, based on values indicated by the local power range monitor sensors and calculated thermal characteristics at the first time and values indicated by the corresponding local power range monitor sensors at the second time, and monitoring the calculated thermal characteristics.