Evidence of design basis accidents mitigation solely with passive safety systems within the frame of the German EASY project


Evidence of design basis accidents mitigation solely with passive safety systems within the frame of the German EASY project

Buchholz, S.; Schaffrath, A.; Bonfigli, G.; Kaczmarkiewicz, N.; Neukam, N.; Schäfer, F.; Wagner, T.

Advanced reactor concepts such as Generation III, III+ and also SMR provide passive safety systems to cope with design basis accidents like loss of coolant accident or loss of main heat sink. In order to be able to assess the controllability of such DBA with passive systems only, computer codes are needed which are able to simulate the behaviour of these passive systems and which are well validated. In the German EASY project the coupled code system AC2 –mainly composed of the system code ATHLET-(CD) and the containment code COCOSYS –is currently being enhanced and validated for such applications on the basis of the KERENA reactor concept of AREVA.
Beyond the implementation of a suitable and effective coupling of the two codes ATHLET and COCOSYS, code development is done for modelling the behaviour of a passive flooding valve and enhancing the ATHLET 3D-model for large water pools.
Validation of the codes is done in two ways: Firstly, single effect tests, performed at the INKA facility in Karlstein, are used to validate the codes for the passive components used in the KERENA design. Secondly, the coupled code system is validated by simulating several integral tests which will be performed at the INKA facility during EASY. These integral tests represent design basis accidents such as SB-LOCA, LB-LOCA and SBO.
Finally, an uncertainty and sensitivity analyses of two integral tests will be performed since the behaviour of passive safety systems is affected strongly by the respective boundary conditions of the system.

Keywords: EASY; Passive Safety Systems; Validation; Code Coupling; AC2

  • Contribution to proceedings
    17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), 03.-08.09.2017, Xi’an, China
    Proceedings of NURETH-17
  • Lecture (Conference)
    17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), 03.-08.09.2017, Xi’an, China

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Publ.-Id: 25346