Analysis of a Simulated Small Break in the Semiscale System under Loss-of-Coolant Accident Conditions

摘要

The Semiscale Mod-1 experimental program conducted by EG and G Idaho, Inc., is part of the overall U.S. Nuclear Regulatory Commission (NRC) and Department of Energy (DOE) sponsored research and development program to investigate the behavior of the pressurized water reactor (PWR) system during an hypothesized loss-of-coolant accident (LOCA). The Semiscale Mod-1 program is intended to provide transient thermal-hydraulic data from a simulated LOCA using a small-scale experimental nonnuclear system. The Semiscale Mod-1 program is a major contributor of experimental data that provide a means of evaluating the adequacy of overall system analytical models as well as the models of the individual system components. Selected experimental data produced by this program will also be used to aid other DOE and NRC sponsored experimental programs, such as the Loss-of-Fluid Test (LOFT) program in optimizing test series, selecting test parameters, and evaluating test results. The Semiscale Mod-1 tests are performed with an experimental system which simulates the principal features of a nuclear plant but which is smaller in volume. Nuclear heating is simulated in the tests by a core composed of an array of electrically heated rods. The core is contained in a pressure vessel which also includes a downcomer, lower plenum, and upper plenum. The Semiscale system piping is arranged such that the intact loop represents three loops of a four-loop nuclear plant, and the broken loop represents the fourth loop. In the present configuration the intact loop contains an active steam generator and pump, and the broken loop contains passive simulators for the steam generator and pump. (ERA citation 04:008267)