REPP, a digital computer method for designing pressure water and boiling water reactor cores within specified heat transfer and fuel centerline temperature limits is presented. The method incorporates the Westinghouse W-2 and W-3 empirical correlations and a theoretical hot channel model to predict burnout conditions in a rod bundle. Two geometries are considered; rods in a triangular array and rods in a square lattice. The heat transfer problem solved is a one-dimensional analysis. Pressure drop is considered for four types of fuel-pin spacers. Variable heat generation rate through the fuel-pin and sintering in low density fuels are also included.

Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain the understanding of the conceptual design elements and specific analysis methods of supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference to engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology.

This document presents an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), and presentation and investigation of CHF data acquired at the University of Wisconsin --- Madison. The HPCHF facility has been designed and built to study CHF at moderate to high pressures and low mass flux ranges in a square rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. Critical heat flux is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (i.e. moderate to high pressure and low mass fluxes). Presently available open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. The HPCHF facility test section is a heated bundle comprised of four electrically heated elements arranged in a 2x2 rectangular lattice with a prototypic chopped-cosine axial power profile. Thermocouples at various axial and circumferential positions embedded in each heater monitor for CHF occurrence. The inlet parameter ranges under investigation are: subcoolings of 250 --- 650 kJ/kg, pressures of 8 --- 16 MPa, and mass fluxes of 500 --- 1600 kg/m2s. The gathered data have been analyzed for primary parametric parameter dependency and compared to a selection of existing CHF prediction methods whose application ranges may be applicable to the conditions of SMRs. Furthermore, an Artificial Neural Network (ANN) methodology has been applied to this studies database in order to overcome some of the dataset's limitations and provide additional insight.