Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

Abstract In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant. The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge. This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.

[1]  Marie-France Robbe,et al.  Simulation of the depressurisation occuring at the beginning of a LOCA in a 4-loop PWR , 2003 .

[2]  Marie-France Robbe,et al.  A porosity method to describe the influence of internal structures on a fluid flow in case of fast dynamics problems , 2002 .

[3]  S. H. Fistedis,et al.  Analysis of the Primary Containment Response using a Hydrodynamic-elastic-plastic Computer Code , 1974 .

[4]  Marie-France Robbe,et al.  Hydrodynamic loads on a PWR primary circuit due to a LOCA , 2002 .

[5]  M. S. Cowler,et al.  Dynamic Fluid-Structure Analysis of Shells Using the PISCES 2 DELK Computer Code , 1979 .

[6]  M. Lepareux,et al.  PLEXUS - a General Computer Program for the Fast Dynamic Analysis - the Case of Pipe-Circuits , 1985 .

[7]  H. Holtbecker.  Testing philosophy and simulation techniques , 1977 .

[8]  A. L. Florence,et al.  Scale modelling in LMFBR safety , 1979 .

[9]  B. L. Smith,et al.  Status of Coupled Fluid-Structure Dynamics Code SEURBNUK , 1983 .

[10]  Y. W. Chang,et al.  Application of containment codes to LMFBRs in the United States , 1977 .

[11]  Pascal Galon,et al.  Hydrogen combustion loads—plexus calculations , 1997 .

[12]  J. L. Graveleau,et al.  Calculation of Fluid-Structure Interaction for Reactor Safety with the Cassiopee Code , 1979 .

[13]  A. Benuzzi,et al.  The Computer Code SEURBNUK-2 for Fast Reactor Explosion Containment Safety Studies , 1977 .

[14]  Marie-France Robbe,et al.  Modeling of the depressurisation induced by a pipe rupture in the primary circuit of a nuclear plant , 2003 .

[15]  C. Fiche,et al.  A Code Comparison Exercise Based on the LMFBR Containment Experiment MARA-04 , 1985 .

[16]  N. E. Hoskin,et al.  The cova programme for the validation of computer codes for fast reactor containment studies , 1978 .

[17]  M. Lepareux,et al.  Numerical simulation of an explosion in a simple scale model of a nuclear reactor , 2002 .

[18]  Y. Cariou,et al.  LMR large accident analysis method , 1997 .

[19]  A. Benuzzi,et al.  Comparison of Different LMFBR Primary Containment Codes Applied to a Benchmark Problem , 1987 .

[20]  P. Verpeaux,et al.  PLEXUS: A General Computer Code for Explicit Lagrangian Computation , 1979 .

[21]  M. F. Robbe,et al.  Numerical interpretation of the mara 8 experiment simulating a hypothetical core disruptive accident , 2003 .

[22]  F. Casadei,et al.  Use of PLEXUS as a LMFBR Primary Containment Code for the CONT Benchmark Problem , 1989 .

[23]  A. Benuzzi,et al.  The COVA program—Validation of the fast reactor containment code seurbnuk☆ , 1980 .

[24]  M. Cigarini,et al.  Applications of ASTARTE-4 Code to Explosive Models with Complex Internal Structure Using the Rezoning Facility , 1983 .