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ABWR Start-Up Test Analysis With Transient Code BISON

[+] Author Affiliations
Kazuki Yano, Yamato Hayashi

TOSHIBA Corporation, Yokohama, Kanagawa, Japan

Paper No. ICONE17-75127, pp. 267-273; 7 pages
doi:10.1115/ICONE17-75127
From:
  • 17th International Conference on Nuclear Engineering
  • Volume 5: Fuel Cycle and High and Low Level Waste Management and Decommissioning; Computational Fluid Dynamics (CFD), Neutronics Methods and Coupled Codes; Instrumentation and Control
  • Brussels, Belgium, July 12–16, 2009
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4355-0 | eISBN: 978-0-7918-3852-5
  • Copyright © 2009 by ASME

abstract

New ABWR plants are planned or studied in Japan as well as the U.S. and other countries. Multiple safety analyses have been performed for ABWR in order to demonstrate the safe and stable performance of this type of plants. Westinghouse code package (BISON, PHOENIX4 and POLCA7) has been used for most BWR plants in Europe and the U.S., while it has not been used for ABWR transient analyses yet. In order to verify the applicability of BISON and POLCA7 code to ABWR, input data for these codes have been generated in accordance with geometry and detailed design specifications, such as control system response and set points of various transient mitigation functions. The calculated steady-state and transient results are compared to the recorded test data of past ABWR start-up tests performed prior to commercial operation. For the validation of BISON code, control system tests (pressure, recirculation flow, feedwater flow), feedwater pump trip, reactor internal pump trip, load rejection and MSIV closure tests have been simulated and the obtained results show an overall good agreement with actual plant behavior. This validation analysis used output from POLCA7 as the input of kinetics with verified calculation for core performance of steady-state. Westinghouse code package was validated for applying to ABWR by compared with the data collected from startup test. It is expected to be capable for simulating transients of the reactor, therefore, it is adequate enough to design and predict of transient behavior for ABWR.

Copyright © 2009 by ASME

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