LMFBR Safety and Sodium Boiling PDF Download
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Languages : en
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Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.
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Languages : en
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Book Description
Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.
![Comparison of One- and Two-dimensional Sodium-boiling Models. [LMFBR]. PDF](https://books.google.com/books/content/images/frontcover/95k6nQAACAAJ?fife=w80-h100)
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Languages : en
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Prediction of sodium boiling and dryout in the fuel subassemblies of a Liquid Metal Fast Breeder Reactor (LMFBR) under certain accident aconditions is of paramount importance in LMFBR safety. In the present paper, boiling and dryout calculations performed by two computer codes, one-dimensional (1-D), LOOP-1 and two-dimensional (2-D), THORAX are presented and compared with experimental data.
Author: Songbai Cheng
Publisher: Springer Nature
ISBN: 9811661162
Category : Science
Languages : en
Pages : 313
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Book Description
This book highlights the advances and trends in the safety analysis of sodium-cooled fast reactors, especially from the perspective of particle bed-related phenomena during core disruptive accidents. A sodium-cooled fast reactor (SFR) is an optimized candidate of the next-generation nuclear reactor systems. Its safety is a critical issue during its R&D process. The book elaborates on research progresses in particle bed-related phenomena in terms of the molten-pool mobility, the molten-pool sloshing motion, the debris bed formation behavior, and the debris bed self-leveling behavior. The book serves as a good reference for researchers, professionals, and postgraduate students interested in sodium-cooled fast reactors. Knowledge provided is also useful for those who are engaging in severe accident analysis for lead-cooled fast reactors and light water reactors.
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Languages : en
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Loss-of-flow (LOF) accidents are of major importance in LMFBR safety. Tests have been performed to simulate the simultaneous failure of all primary pumps and reactor shutdown systems in a 37-pin electrically heated test bundle installed in the KNS sodium boiling loop at the Institute of Reactor Development, Karlsruhe. The tests simulated LOF conditions of the German prototype LMFBR, the SNR 300. The main objectives of these tests were to characterize the transient boiling development to cladding dryout and to provide data for validation of sodium boiling codes. One particular LOF test, designated L22, at full power was selected as a benchmark exercise for comparison of several codes at the Eleventh Meeting of the Liquid Metal Boiling Working Group (LMBWG) held in Grenoble, France, in October 1984. In this paper, the results of the calculations performed at ORNL with the two-dimensional (2-D) boiling code THORAX are presented.
Author: Rockwell International. Atomics International Division
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 62
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Languages : en
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The objectives of this project are to: develop a computer program for calculating two-dimensional, transient, natural convection phenomena such as those arising from various sodium spill accidents in Liquid Metal Fast Breeder Reactor (LMFBR) heat transfer equipment vaults, head compartments, containment buildings, and secondary heat transfer systems; develop experimental programs and conduct tests that will characterize the behavior of sodium, sodium oxide, fuel, fission product, and other aerosols as they might be generated by various postulated LMFBR accidents; determine by analysis and experiment the generation and transport of these aerosols; and determine the effects of an accident in an LMFBR involving fuel melting by contacting molten UO2 (a fuel simulant) with stainless steel, sodium, concrete, and various sacrificial materials.
Author: N. Sheriff
Publisher:
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Category :
Languages : en
Pages : 20
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Category : Nuclear energy
Languages : en
Pages : 680
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Languages : en
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Objective is to improve current understanding of sodium voiding behavior under postulated LMFBR accident conditions. Multi-dimensional computer models are being developed under low flow and low power conditions. The following computer codes are being developed and assessed: NATOF-2D, THERMIT-S-6E, and THERMIT-S-4E. The effect of virtual mass on the characteristics and numerical stability in two-phase flows was studied. (DLC).
![Post-test Analysis of Dryout Test 7B' of the W-1 Sodium Loop Safety Facility Experiment with the SABRE-2P Code. [LMFBR]. PDF](https://books.google.com/books/content/images/frontcover/UwZXAQAACAAJ?fife=w80-h100)
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Languages : en
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An understanding of conditions that may cause sodium boiling and boiling propagation that may lead to dryout and fuel failure is crucial in liquid-metal fast-breeder reactor safety. In this study, the SABRE-2P subchannel analysis code has been used to analyze the ultimate transient of the in-core W-1 Sodium Loop Safety Facility experiment. This code has a 3-D simple nondynamic boiling model which is able to predict the flow instability which caused dryout. In other analyses dryout has been predicted for out-of-core test bundles and so this study provides additional confirmation of the model.
