Dynamic Behavior of Boiling Water Reactors

A study of the basic processes involved in boiling water nuclear reactor dynamics is presented. The main emphasis of this research has been placed on the physical interpretation of these processes. It is shown that this type of reactors have two regimes of operation: linear, during normal operation, and nonlinear, if they become unstable due to the thermohydraulic feedback. Both of these regimes are studied using low-order physical models. The main result obtained from the linear study is the pole-zero configuration of the reactivity-to-power transfer function. It is determined that three zeros and four poles are needed to properly represent this transfer function. Physical processes are identified with these transfer function features. Based on the understanding of these processes, an automated algorithm to estimate boiling water reactor stability from neutron noise measurements is developed and implemented as a computer code. The causative mechanism leading to the appearance of the limit cycle in boiling water reactors is identified from the nonlinear study. The relationship between the different process variables during limit cycle oscillations is studied. It is shown that these oscillations could reach large amplitudes. The stability of the limit cycle is also studied. It is shown that the amplitude of the limit cycle can become unstable and produce period-doubling pitchfork bifurcations which scale according to Feigenbaum\u27s universality theory. As a consequence of the bifurcation process, aperiodic solutions of the deterministic reactor equations are found to be possible. Finally, nonlinear noise propagation is studied. A nonperturbative technique is developed for detecting the onset of linear instability and the transition to the nonlinear regime

NERI Project 99-119. A New Paradigm for Automatic Development of Highly Reliable Control Architectures for Nuclear Power Plants. Phase-2 Progress Report

This report describes the tasks performed and the progress made during Phase 2 of the DOE-NERI project number 99-119 entitled Automatic Development of Highly Reliable Control Architecture for Future Nuclear Power Plants. This project is a collaboration effort between the Oak Ridge National Laboratory (ORNL), The University of Tennessee, Knoxville (UTK) and the North Carolina State University (NCSU). ORNL is the lead organization and is responsible for the coordination and integration of all work

NERI PROJECT 99-119."A NEW PARADIGM FOR AUTOMATIC DEVELOPMENT OF HIGHLY RELIABLE CONTROL ARCHITECTURES FOR NUCLEAR POWER PLANTS."PHASE-1 PROGRESS REPORT

This report describes the tasks performed and the progress made during Phase 1 of the DOE-NERI project number 99-119 entitled ''Automatic Development of Highly Reliable Control Architecture for Future Nuclear Power Plants''. This project is a collaboration effort between the Oak Ridge National Laboratory (ORNL,) The University of Tennessee, Knoxville (UTK) and the North Carolina State University (NCSU). ORNL is the lead organization and is responsible for the coordination and integration of all work. This research focuses on the development of methods for automated generation of control systems that can be traced directly to the design requirements for the life of the plant. Our final goal is to ''capture'' the design requirements inside a ''control engine'' during the design phase. This control engine is, then, not only capable of designing automatically the initial implementation of the control system, but it also can confirm that the original design requirements are still met during the life of the plant as conditions change. This control engine captures the high-level requirements and stress factors that the control system must survive (e.g. a list of transients, or a requirement to withstand a single failure). The control engine, then, is able to generate automatically the control-system algorithms and parameters that optimize a design goal and satisfy all requirements. As conditions change during the life of the plant (e.g. component degradation, or subsystem failures) the control engine automatically ''flags'' that a requirement is not satisfied, and it can even suggest a modified configuration that would satisfy it. This control engine concept is shown schematically in Fig. 1. The implementation of this ''control-engine'' design methodology requires the following steps, which are described in detail in the attachments to this report: (1) Selection of Design Requirements Related to Control System Performance; (2) Implementation of Requirements in Mathematical Form; (3) Development of a Control Algorithm Library; (4) Development and Validation of Plant Models; (5) Automated Control Design Development; (6) Development of Control Architectures; (7) Control Design Implementation; and (8) Development of Diagnostics Methods to Update the Plant Model

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

Lenguaje visual y animación 3d. Propuesta educativa de desarrollo de la alfabetización visual para el disfrute del producto 3d

El objetivo de este trabajo es realizar una investigación sobre las aplicaciones del uso de la luz y del color como recurso comunicativo, narrativo y expresivo en el caso concreto de la animación 3d. Para ello se propone una herramienta de análisis que se aplica, en un trabajo de campo, sobre secuencias seleccionadas de productos de PIXAR Animation Studios. Los resustados obtenidos ofrecen datos sobre el empleo de la luz y del color como recursos comunicativos, narrativos y expresivos en las muestras seleccionadas. Por último, del análisis de estos datos, se concluye la importancia y capacidad de uso de los recursos estudiados para la optimización de la comunicación, narración y expresión en el cine de animación 3d. Por otro lado, la herramienta propuesta aporta un método para la observación reflexiva y crítica de la imagen dirigida al estudio de recursos, técnicas y estrategias visuales. La aplicación de la herramienta proporciona ejemplos prácticos y aplicados de los recursso visuales estudiados: la luz y el color. Por ello, tanto la herramienta como los análisis realizados, se convierten en una propuesta para el desarrollo de la alfabetización visual aplicada a la animación 3d, campo que, actualmente, se encuentra en plena expansión. La necesidad e importancia de la alfabetización visual es recogida en diferentes informes y declaraciones elaborados por el Ministerio de Educación y Ciencia, Organismos como la UNESCO así como por investigadores reconocidos internacionalmente. se considera que en un mundo cada vez más dominado por las imágenes, tanto los creadores como los espectadores deben disponer de unos conocimientos de lenguaje visual que les permitan valorar tanto el mensaje como la calidad de los productos audiovisuales que crean o que les rodean. Según los citados informes y autores promocionar la alfabetización visual en un mundo dominado por lo visual permitirá, entre otros, hacer más eficaz la comunicación, formentar la creatividad, la individualidad, la observación crítica y el disfrute visualMarch Leuba, ME. (2011). Lenguaje visual y animación 3d. Propuesta educativa de desarrollo de la alfabetización visual para el disfrute del producto 3d [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/10742Palanci

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

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

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

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

Implementation of the Fissile Mass Flow Monitor Source Verification and Confirmation

This report presents the verification procedure for neutron sources installed in U.S. Department of Energy equipment used to measure fissile material flow. The Fissile Mass Flow Monitor (FMFM) equipment determines the {sup 235}U fissile mass flow of UF{sub 6} gas streams by using {sup 252}Cf neutron sources for fission activation of the UF{sub 6} gas and by measuring the fission products in the flow. The {sup 252}Cf sources in each FMFM are typically replaced every 2 to 3 years due to their relatively short half-life ({approx} 2.65 years). During installation of the new FMFM sources, the source identity and neutronic characteristics provided by the manufacturer are verified with the following equipment: (1) a remote-control video television (RCTV) camera monitoring system is used to confirm the source identity, and (2) a neutron detection system (NDS) is used for source-strength confirmation. Use of the RCTV and NDS permits remote monitoring of the source replacement process and eliminates unnecessary radiation exposure. The RCTV, NDS, and the confirmation process are described in detail in this report