An Experimental and Modelling Study of Natural-circulation Boiling Water Reactor Dynamics


Book Description

Contents of this Doctoral Dissertation include: Understanding the linear stability characteristics of BWRs, Experiments on the stability of the Desire facility, Applications of the reducer-order model, Numerical analysis of the nonlinear dynamics of BWRs, Experiments on the nonlinear dynamics of natural-circulation two-phase flows, Experiments on the neutronic-thermalhydraulic stability, Conclusions and Discussion




Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs


Book Description

20% of the Nuclear Power Plants are known as Boiling Water Reactors (BWRs). These BWRs have pumps that cool their reactor. In the design of new BWRs, ways to cool the core by a natural circulation flow, without pumps, also called natural circulation BWRs, are being considered. In these new systems, a chimney is installed on top of the core to increase natural circulation flow. A possible disadvantage of natural circulation BWRs might be their susceptibility to instabilities, which could then lead to both flow and power oscillations. The stability features of both natural circulation and forced circulation BWRs have been investigated thoroughly, using dedicated experimental setups, analytical models and numerical codes. We distinguish between pure thermal-hydraulic stability - where the fission power is assumed to be constant - and coupled thermalhydraulic-neutronic stability - where the two-phase mixture in the core influences the fission chain reaction...




Experimental and Numerical Stability Investigations on Natural Circulation Boiling Water Reactors


Book Description

In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs such as the natural circulation boiling water reactor (BWR). In such a reactor, however, the flow is not a controlled parameter but is dependent on the power. As a result, the dynamical behavior significantly differs from that in conventional forced circulation BWRs. For that reason, predicting the stability characteristics of these reactors has to be carefully studied. In this work, a number of open issues are investigated regarding the stability of natural circulation BWRs (e.g. margins to instabilities at rated conditions, interaction between the thermal-hydraulics and the neutronics, and the occurrence of flashing induced instabilities) with a strong emphasis on experimental evidence.




Computational Fluid Dynamics Analysis of Natural Circulation Flows in a Pressurized-Water Reactor Loop Under Severe Accident Conditions


Book Description

Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the steam generator of a pressurized-water reactor (PWR) during a severe accident scenario. The results extend earlier predictions of steam generator inlet plenum mixing with the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the steam generator. Tube leakage and mass flow into the pressurizer surge line are also considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg too. The new model is consistent with the predicted behavior and accounts for flow into a side mounted surge line if present. A density- based Froude number correlation is utilized to provide a method for determining the flow rate from the vessel to the hot leg directly from the conditions at the ends of the hot leg pipe. This provides a physically based approach for establishing the hot leg flows. The mixing parameters and correlations are proposed as a best-estimate approach for estimating the flow rates and mixing in one-dimensional system codes applied to severe accident natural circulation conditions. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the steam generator secondary side temperatures or heat transfer rates to the steam generator. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. A further modeling improvement is suggested regarding the application of thermal entrance effects in the hot leg and surge line. This work supports the U.S. Nuclear Regulatory Commission studies of steam generator tube integrity under severe accident conditions.







An Analytical and Experimental Investigation of Natural Circulation Transients in a Model Pressurized Water Reactor


Book Description

Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS), '' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.




Natural Circulation Loop Performance at 1000 Psia Under Periodic Accelerations


Book Description

Initial experimental and analog results have been obtained from a study of the effect of shipboard disturbances on the heat transfer and fluid flow performance of boiling water reactors. Of special interest is the influence of ship's motion on the fuel element burnout level and the hydraulic stability of the reactor.




Dynamic Analysis of Coolant Circulation in Boiling Water Nuclear Reactors


Book Description

The dynamics of two-phase flow through the coolant channels of a natural-circulation boiling water nuclear reactor is studied analytically. One-dimensional conservation equations describing the flow through each channel are written in a linearized perturbed form, and Laplace transformation in time is performed. A systematic procedure is developed to approximate the solution. The solution may be oscillatory both in time and space, and the stability depends largely upon the steady-state profile of velocity and void fraction along the channel, as well as the channel length. The simplifying assumption made by earlier investigators that the slip ratio is constant along the channel length is shown to yield results close to the true solution.







Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors


Book Description

Based on an IAEA coordinated research project focused on the use of passive safety systems and natural circulation to help meet the safety and economic goals of advanced nuclear power plants, this publication includes the identification and definition of the thermo-hydraulic phenomena that affect the reliability of passive safety systems, characterization of each phenomenon, integral tests to examine the passive systems and natural circulation, and a methodology for examining passive system reliability.