An Experimental and Modelling Study of Natural-circulation Boiling Water Reactor Dynamics - read free eBook in online reader or directly download on the web page. Select files or add your book in reader. Download and read online ebook An Experimental and Modelling Study of Natural-circulation Boiling Water Reactor Dynamics write by Róbert Zboray. This book was released on 2002. An Experimental and Modelling Study of Natural-circulation Boiling Water Reactor Dynamics available in PDF, EPUB and Kindle. 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
Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs - read free eBook in online reader or directly download on the web page. Select files or add your book in reader. Download and read online ebook Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs write by Masahiro Furuya. This book was released on 2006. Experimental and Analytical Modeling of Natural Circulation and Forced Circulation BWRs available in PDF, EPUB and Kindle. 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
Experimental and Numerical Stability Investigations on Natural Circulation Boiling Water Reactors - read free eBook in online reader or directly download on the web page. Select files or add your book in reader. Download and read online ebook Experimental and Numerical Stability Investigations on Natural Circulation Boiling Water Reactors write by Christian Pablo Marcel. This book was released on 2007. Experimental and Numerical Stability Investigations on Natural Circulation Boiling Water Reactors available in PDF, EPUB and Kindle. 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.
Analyses of Experimental Power-reactivity Feedback Transfer Functions for a Natural Circulation Boiling Water Reactor
Analyses of Experimental Power-reactivity Feedback Transfer Functions for a Natural Circulation Boiling Water Reactor - read free eBook in online reader or directly download on the web page. Select files or add your book in reader. Download and read online ebook Analyses of Experimental Power-reactivity Feedback Transfer Functions for a Natural Circulation Boiling Water Reactor write by J. A. DeShong. This book was released on 1958. Analyses of Experimental Power-reactivity Feedback Transfer Functions for a Natural Circulation Boiling Water Reactor available in PDF, EPUB and Kindle. The use of the simplified model for parameter studies is demonstrated by a series of calculations to evaluate the effect of heat transfer time constant on stability.
Computational Fluid Dynamics Analysis of Natural Circulation Flows in a Pressurized-Water Reactor Loop Under Severe Accident Conditions
Computational Fluid Dynamics Analysis of Natural Circulation Flows in a Pressurized-Water Reactor Loop Under Severe Accident Conditions - read free eBook in online reader or directly download on the web page. Select files or add your book in reader. Download and read online ebook Computational Fluid Dynamics Analysis of Natural Circulation Flows in a Pressurized-Water Reactor Loop Under Severe Accident Conditions write by U.s. Nuclear Regulation Comion. This book was released on 2014-06-30. Computational Fluid Dynamics Analysis of Natural Circulation Flows in a Pressurized-Water Reactor Loop Under Severe Accident Conditions available in PDF, EPUB and Kindle. 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.
