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An Experimental Investigation Into Potential Nuclear Reactor Vessel Flow Instabilities

[+] Author Affiliations
Jeffrey A. Brown

Arizona State University, Tempe, AZ

James W. Rowland

Arizona Public Service, Tonopah, AZ

H. Joseph Fernando

University of Notre Dame, Notre Dame, IN

Paper No. ICONE18-30211, pp. 1177-1184; 8 pages
doi:10.1115/ICONE18-30211
From:
  • 18th International Conference on Nuclear Engineering
  • 18th International Conference on Nuclear Engineering: Volume 4, Parts A and B
  • Xi’an, China, May 17–21, 2010
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4932-3
  • Copyright © 2010 by ASME

abstract

An investigation into the increase in Plant Protection System (PPS) alarms at a three-unit US Pressurized Water Reactor (PWR) plant has determined that the alarms are the result, in part, of a hydraulic instability that has developed within the Reactor Coolant System (RCS) following the replacement of the steam generators in all three units of the Palo Verde Nuclear Generating Station (PVNGS). An experimental effort has been established by Arizona Public Service Company and Arizona State University in an attempt to determine the cause of these instabilities. Preliminary investigations have determined that the time scale of these instabilities is consistent with larger scale transient flow processes of the reactor vessel. Accordingly, the flow characteristics were assessed and localized flow measurements made using a one-fifth scale physical model of the upper plenum region of the reactor core of the Combustion Engineering System 80 reactor vessel to verify the postulation that large vortex structures referred to as “precessing” vortices [Ref. 1] affect the core exit flow conditions resulting in the noted flow instabilities. The physical model investigation was complemented by numerical analysis based on a Computational Fluid Dynamics (CFD) code performed for the same geometry. Benchmarking of the CFD model by the scaled physical model is intended to provide increased confidence in the CFD code. If verified, the CFD code may be modified so as to establish corrective actions for this condition, where physical modeling would probably be time consuming and cost prohibitive. The initial results for the physical and computational models demonstrate very good agreement between the measured and calculated flows in the upper-plenum region. The results of the complementary experimental and analytic evaluations do not support the presence of any large scale vortices of appropriate space scales that could affect flow conditions within the upper-plenum region. The elimination of the reactor vessel as the source of the instabilities suggests that the replacement steam generators may be the root cause of the flow instabilities. There is a possibility, however, that frequencies pertinent to vortices may be triggering mechanisms for flow instabilities in the entire system.

Copyright © 2010 by ASME

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