Showing papers by "Robert M. Edwards published in 2000"

A best-estimate reactor core monitor using state feedback strategies to reduce uncertainties

Abstract: The development and demonstration of a new algorithm to reduce modeling and state-estimation uncertainty in best-estimate simulation codes has been investigated. Demonstration is given by way of a prototype reactor core monitor. The architecture of this monitor integrates a control-theory-based, distributed-parameter estimation technique into a production-grade best-estimate simulation code. The Kalman Filter-Sequential Least-Squares (KFSLS) parameter estimation algorithm has been extended for application into the computational environment of the best-estimate simulation code RELAP5-3D. In control system terminology, this configuration can be thought of as a 'best-estimate' observer. The application to a distributed-parameter reactor system involves a unique modal model that approximates physical components, such as the reactor, by describing both states and parameters by an orthogonal expansion. The basic KFSLS parameter estimation is used to dynamically refine a spatially varying (distributed) parameter. The application of the distributed-parameter estimator is expected to complement a traditional nonlinear best-estimate simulation code by providing a mechanism for reducing both code input (modeling) and output (state-estimation) uncertainty in complex, distributed-parameter systems.

Stability monitoring tests using a nuclear-coupled boiling channel model

Abstract: A new test platform for stability studies is presented that can be used to generate a power time series, which in turn may be used to validate the capability of boiling water reactor stability-monitoring algorithms. The thermal hydraulics for boiling channels are modeled and coupled with neutron kinetics to analyze the nonlinear dynamics of the closed-loop system, The model uses point kinetics to study core-wide oscillations, and it couples two time-domain calculations, for the fundamental and first harmonic modes, to study out-of-phase oscillations. The channel coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate a noisy power time series. Autoregressive analysis performed with the computer-generated series agrees with the stability properties of the boiling channel. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions.