Flooding Experiments and Modeling for Improved Reactor Safety

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Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line ... continued below

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Solmos, M., Hogan, K.J., VIerow, K. September 14, 2008.

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Description

Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.

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  • Journal Name: Papers will be reviewed for submission to an as yet unspecified journal.; Conference: US-Japan Two Phase Flow Seminar, Santa Monica, CA, Sept. 14-18, 2008, National Science Foundation.

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  • Report No.: DOE/ID/14696-CONF1
  • Grant Number: FG07-05ID14696
  • Office of Scientific & Technical Information Report Number: 936720
  • Archival Resource Key: ark:/67531/metadc897561

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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  • September 14, 2008

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  • Sept. 27, 2016, 1:39 a.m.

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Solmos, M., Hogan, K.J., VIerow, K. Flooding Experiments and Modeling for Improved Reactor Safety, article, September 14, 2008; United States. (digital.library.unt.edu/ark:/67531/metadc897561/: accessed December 16, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.