94 research outputs found
Recommended from our members
Advanced Neutron Source Dynamic Model (ANSDM) code description and user guide
A mathematical model is designed that simulates the dynamic behavior of the Advanced Neutron Source (ANS) reactor. Its main objective is to model important characteristics of the ANS systems as they are being designed, updated, and employed; its primary design goal, to aid in the development of safety and control features. During the simulations the model is also found to aid in making design decisions for thermal-hydraulic systems. Model components, empirical correlations, and model parameters are discussed; sample procedures are also given. Modifications are cited, and significant development and application efforts are noted focusing on examination of instrumentation required during and after accidents to ensure adequate monitoring during transient conditions
Recommended from our members
Importance of momentum dynamics in BWR neutronic stability: experimental evidence
Momentum dynamics affect the boiling water reactor (BWR) neutronic stability by coupling steam void perturbations and core-inlet coolant flow. Computer simulations have shown that proper modeling of the recirculation loop, which shares the upper and lower plena pressures with the reactor core, is essential for accurate stability calculations. Purpose of this paper is to show experimental evidence, obtained from a recent series of stability tests performed at the Browns Ferry-1 BWR, demonstrating the important role of momentum dynamics in BWR neutronic stability
Recommended from our members
Sensitivity of BWR stability calculations to numerical integration techniques
Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters, modeling assumptions, and numerical integration techniques. Following the 1988 LaSalle instability event, a significant industry-wide effort was invested in identifying these sensitivities. One major conclusion from these studies was that existing time-domain codes could best predict BWR stability by using explicit methods for the energy equation with a Courant number as close to unity as possible. This paper presents a series of sensitivity studies using simplified models, which allow us to determine the effect that different numerical integration techniques have on the results of stability calculations. The present study appears to indicate that, even though using explicit integration with a Courant number of one is adequate for existing codes using time-integration steps of less than 10 ms, second-order solution techniques for the time integration can result in significant improvements in the accuracy of linear (i.e., decay ratio) stability calculations
Recommended from our members
Calibration measurements using the ORNL fissile mass flow monitor
This paper presents a demonstration of fissile-mass-flow measurements using the Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor in the Paducah Gaseous Diffusion Plant (PGDP). This Flow Monitor is part of a Blend Down Monitoring System (BDMS) that will be installed in at least two Russian Federation (R.F.) blending facilities. The key objectives of the demonstration of the ORNL Flow Monitor are two: (a) demonstrate that the ORNL Flow Monitor equipment is capable of reliably monitoring the mass flow rate of {sup 235}UF{sub 6} gas, and (b) provide a demonstration of ORNL Flow Monitor system in operation with UF{sub 6} flow for a visiting R.F. delegation. These two objectives have been met by the PGDP demonstration, as presented in this paper
Recommended from our members
Excitation sources for fuel assembly vibrations in a PWR
Noise measurements from in-core neutron detectors have been utilized previously to monitor in-plant vibrations of pressurized water reactor (PWR) fuel assemblies. Fuel assembly resonant frequencies and mode shapes were obtained from these in-core measurements by observing resonances in the neutron noise power spectral density (PSD) and then plotting the axial dependence of the root mean square (rms) neutron noise over frequency ranges containing these resonances. In order to determine the fuel assembly mode shapes and their relationship to core barrel motion, we performed simultaneous measurements of in-core and ex-core neutron noise at the Sequoyah-1 reactor, a Westinghouse 1150 MW(e) PWR. Analysis of this data indicates that there are two different sources of vibrational excitation for the fuel in the 6- to 8-Hz frequency range. 7 refs., 2 figs
Recommended from our members
Passive Nmis Measurements to Estimate Shape of Plutonium Assemblies (Slide Presentation)
The purpose of this work is to estimate shape of plutonium assemblies using new signatures acquired by passive NMIS measurements (no external source). Applications include identification of containerized regular shapes of plutonium, identification by shape without template, verification of shape for template initialization, and potential utility for estimating shape of holdup in plutonium processing facilities. To illustrate the technique and test its feasibility, laboratory measurements have been performed with californium spontaneous fission sources as a surrogate for plutonium. Advantages of the technique include the following: passive (requires no external source for plutonium measurements), stationary (no scanning of the assembly is required), penetrative (shape is estimated from neutron emissions), obscurable (spatial resolution can be deliberately degraded by changing detector size and/or timing resolution), inexpensive (majority of NMIS components are commercial products), portable (detection system is transported to the item, not vice versa). It is concluded that passive NMIS measurements can infer the mass of plutonium assemblies: NMIS correlations scale directly with spontaneous fission rate (Pu-240); NMIS correlations scale with fissile mass (Pu-239) and multiplication. New third-order correlations can estimate the shape of fission sources (Pu-240 & Pu-239) from passive measurements. Surrogate measurements of californium spontaneous fission sources have demonstrated the feasibility of this concept. Measurements of various shapes of plutonium are necessary to continue the development of this technique
Recommended from our members
Optimal filtering, parameter tracking, and control of nonlinear nuclear reactors
This paper presents a new formulation of a class of nonlinear optimal control problems in which the system's signals are noisy and some system parameters are changing arbitrarily with time. The methodology is validated with an application to a nonlinear nuclear reactor model. A variational technique based on Pontryagin's Maximum Principle is used to filter the noisy signals, estimate the time-varying parameters, and calculate the optimal controls. The reformulation of the variational technique as an initial value problem allows this microprocessor-based algorithm to perform on-line filtering, parameter tracking, and control
Recommended from our members
Understanding the boiling water reactor limit cycle
This paper presents an interpretation of the physical mechanisms involved in the development of limit cycle oscillations in boiling water reactors (BWRs). Based on this interpretation, approximate correlations for some oscillation parameters are developed and shown to be largely independent of the particular reactor operating condition. The stability of the limit cycle is also studied in this paper. It is shown that the BWR limit cycle may become unstable and bifurcate. The bifurcation process leads to aperiodic (chaotic) behavior of the reactor power and causes the peak oscillation powers to be larger than those from a nonbifurcated limit cycle. 7 refs., 6 figs., 1 tab
Recommended from our members
A new approach to controlling the water level of U-tube steam generators
Automatic water level control in steam generators is currently achieved via a three-element controller. This algorithm is based on the measurements of level, steam flow, and feedwater flow. Unfortunately, at low power the feedwater flow signal is highly unreliable, forcing the transfer to manual control. A large number of reactor trips occur under these conditions, and the nuclear industry has shown a concern for this problem. This paper proposes and validates an alternative automatic control algorithm. This new algorithm does not rely on flow signals; it uses instead the pressure measurement in the steam header. A level set point modulation is introduced that allows the algorithm to compensate for shrink and swell phenomena. The standard {Delta}-P algorithm to control feedwater pump speed also has been modified to achieve improved performance and integration. 2 refs., 12 figs
- …