1,721,003 research outputs found
A Numerical and Experimental Investigation of the Single-Phase Natural Circulation with Multiloop
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Conceptual Design of an Inherently Safe and Simple Tube Reactor Using Water as Moderator and Coolant
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An independent assessment of Groeneveld et al.s 1995 CHF look-up table
No full textThe prediction capability of the 1995 CHF look-up table (Groeneveld D.C., et al., Nucl. Eng. Des. 163 (1996) 1-23) is independently assessed based on the KAIST data base consisting of 10 822 data for uniformly-heated, vertical, round tubes. This confirms the error statistics for the heat balance method reported by Groeneveld et al. and shows overall average and RMS errors of 4.2 and 36.7%, respectively, for the direct substitution method. The new 1995 table shows better prediction capability than the 1986 AECL-UO table (Groeneveld et al., 1986), especially for the low-pressure, low-flow region. The error analysis indicates the length effect even for significantly long tubes. (C) 1997 Elsevier Science S.A
DEVELOPMENT OF A BUNDLE CORRECTION METHOD AND ITS APPLICATION TO PREDICTING CHF IN ROD BUNDLES
No full textA bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions
Assessment of passive containment cooling concepts for advanced pressurized water reactors
No full textThe containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants, therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. In addition to the conventional active systems, several passive containment cooling system (PCCS) concepts are developed or proposed for advanced pressurized water reactors (PWRs) to ensure the operability without the need for electric power. This paper presents an overview and comparative assessment on selected PCCS concepts for advanced PWRs for both steel and concrete containment systems. Major concepts considered are: (a) the spray of water on the outer surface of steel containment from an elevated tank, (b) an external moat for steel containment, (c) a suppression pool for concrete containment, and (d) combination of the internal spray and internal or external condensers for concrete containment. Emphasis is given to the heat removal principles, required heat transfer area, system complexity and operational reliability. (C) 1997 Elsevier Science Ltd
Improved methodology for generation of axial flux shapes in digital core protection systems
No full textAn improved method of axial flux shape (AFS) generation for digital core protection systems of pressurized water reactors is presented in this paper using an artificial neural network (ANN) technique-a feedforward network trained by backpropagation. It generates 20-node axial power shapes based on the information from three ex-core detectors. In developing the method, a total of 7173 axial flux shapes are generated from ROCS code simulation for training and testing of the ANN. The ANN trained 200 data predicts the remaining data with the average root mean square error of about 3%. The developed method is also tested with the real plant data measured during normal operation of Yonggwang Unit 4. The RMS errors in the range of 0.9 similar to 2.1% are about twice as accurate as the cubic spline approximation method currently used in the plant. The developed method would contribute to solve the drawback of the current method as it shows reasonable accuracy over wide range of core conditions. (C) 2002 Elsevier Science Ltd. All rights reserved
Visualization of the subcooled flow boiling of R-134a in a vertical rectangular channel with an electrically heated wall
No full textA digital photographic study of subcooled flow boiling with R-134a was performed in a vertical rectangular channel heated from one side to observe near-wall structures. Primary attention was given to bubble formation and bubble coalescence in the bubble layer. The visualization made possible a detailed observation of the characteristic near-wall region, consisting of vapor remnants, an interleaved liquid layer, and coalesced adjacent bubbles. In particular, the visualization provided clearer insights about bubble coalescence behaviors during flow nucleate boiling. In addition, it is shown that the near-wall bubble layer of nucleate boiling beneath vapor clots is extinguished and, afterwards, the heated surface is locally covered by large vapor films, at CHF. (C) 2004 Elsevier Ltd. All rights reserved
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