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Transient Validation of Areva’s New Arcadia™ Code System

[+] Author Affiliations
Sebastian Kuch, Mario Leberig, Florian Reiterer, Michael Riedmann

AREVA GmbH, Erlangen, Germany

Richard Brock, Kenneth Rooks

AREVA NP, Inc., Lynchburg, VA

Paper No. ICONE22-30743, pp. V02BT09A019; 6 pages
  • 2014 22nd International Conference on Nuclear Engineering
  • Volume 2B: Thermal Hydraulics
  • Prague, Czech Republic, July 7–11, 2014
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4591-2
  • Copyright © 2014 by ASME


AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers.

ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses.

For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3].

The paper will provide representative example results for the abovementioned validation cases.

Copyright © 2014 by ASME



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