Flow Analysis of a Pressurized Water Reactor Core PDF Download
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Author: Charles Ramon Daily
Publisher:
ISBN:
Category :
Languages : en
Pages : 60
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Book Description
Author: Charles Ramon Daily
Publisher:
ISBN:
Category :
Languages : en
Pages : 60
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Book Description
Author: United Engineers & Constructors, inc
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 218
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Book Description
Author: Charles Francis Bonilla
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 56
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Book Description
Author: Long-sun Tong
Publisher:
ISBN:
Category : Science
Languages : en
Pages : 776
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Book Description
The basic objective of this book is to present the principles underlying the thermal and hydraulic design of pressurized water reactors. In addition, the empirical data, engineering properties, and computer techniques required for design, but not available in conventional handbooks, are presented or referenced. Because of the many advances and changes that have occurred since the second edition, extensive improvements in both understanding the phenomena involved and in calculational techniques are reflected in the substantial additions to this third edition. Also, an additional chapter has been added to accommodate the many developments in the area of safety analysis.
Author: John Gittus
Publisher:
ISBN: 9780854033065
Category : Pressurized water reactors
Languages : en
Pages : 19
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Book Description
Author: P. B. Abramson
Publisher: CRC Press
ISBN: 9780891162629
Category : Technology & Engineering
Languages : en
Pages : 418
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Book Description
The Guidebook to Light Water Reactor Safety Analysis brings together government and expert researchers entrusted with maintaining the safety of reactors, preventing incidents, and for creating the guidelines for responding appropriately to emergency situations. It includes an overview presented by the U.S. Nuclear Regulatory Commission. One of the most relevant compendiums of its time, it's a volume of both historical and scientific significance and well worth the consideration of those currently involved with maintaining reactor safety..
Author: Joel Weisman
Publisher:
ISBN:
Category : Technology & Engineering
Languages : en
Pages : 136
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Book Description
Author: U.S. Atomic Energy Commission. Division of Reactor Development
Publisher:
ISBN:
Category : Nuclear power plants
Languages : en
Pages : 510
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Book Description
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 256
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Book Description
A thermal hydraulic analysis of the International Reactor Innovative and Secure (IRIS) core has been performed. Thermal margins for steady state and a selection of Loss Of Flow Accidents have been assessed using three methodologies to account for uncertainty. The thermal hydraulic analysis has shown that the IRIS is designed with adequate thermal margin for steady state operation, the locked rotor/shaft shear accident (LR/SS) and for variants of the partial loss of flow accident. To treat uncertainties, three methods were used, ranging from conservative, deterministic methods, to more realistic and computationally demanding Monte Carlo-based methods. To facilitate the computational requirements of the thermal hydraulic analysis, a script- based interface was created for VIPRE. This scripted interface (written in Matlab) supplants the existing file-based interface. This interface allows for repeated, automatic execution of the VIPRE code on a script-modifiable input data, and parses and stores output data to disk. This endows the analyst with much greater power to use VIPRE in parametric studies, or using the Monte Carlo-based uncertainty analysis methodology. The Matlab environment also provides powerful visualization capability that greatly eases the task of data analysis.
Author: R. M. Hiatt
Publisher:
ISBN:
Category : Boiling water reactors
Languages : en
Pages : 196
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Book Description
REPP, a digital computer method for designing pressure water and boiling water reactor cores within specified heat transfer and fuel centerline temperature limits is presented. The method incorporates the Westinghouse W-2 and W-3 empirical correlations and a theoretical hot channel model to predict burnout conditions in a rod bundle. Two geometries are considered; rods in a triangular array and rods in a square lattice. The heat transfer problem solved is a one-dimensional analysis. Pressure drop is considered for four types of fuel-pin spacers. Variable heat generation rate through the fuel-pin and sintering in low density fuels are also included.
