Benchmark Study on Fuel Bundle Degradation in the FPT3 Phebus Test Using the Severe Accident Codes Athlet-CD, ICARE2 and MELCOR

This paper presents the results of pre-test calculations of the PHEBUS FPT3 experiment. The test scenario with the appropriate initial and boundary conditions was provided by the Institut de Radioprotection et de Sûreté Nucléaire. For the analyses, three severe accident codes were used: ATHLET-CD, ICARE2 and MELCOR, respectively. The calculations were focused on the main phenomena occurring in the bundle such as the thermal behaviour, the hydrogen production mainly due to cladding oxidation, the massive degradation of spent fuel and the release of fission products, control rod and structure materials. Using the pre-defined boundary and initial conditions, relatively small deviations between the code results were obtained, which demonstrates that the dominant processes occurring during a severe accident in the core of PWR reactors can be adequately simulated. By applying these codes to a large spectrum of integral tests as well as to plant analyses, one will obtain reliable results on the fuel bundle behaviour. However, the spread in the calculated oxidised boron carbide masses indicates that modelling efforts are still necessary in all the codes in this respect.JRC.F.4-Nuclear design safet

Reassessment of Computational Tools for the Modeling of Heat Transfer in a Molten UO2 Pool in Natural Convection

International audienceAbstract This work, performed within the European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) H2020 European project, is part of a larger framework intending to reassess the modeling of heat transfer in molten pools on SCARABEE available experimental results. This paper presents simulation results of the in-pile Bain Fondu (BF1) test, performed within the SCARABEE-N program, using accident source term evaluation code (ASTEC), Sn method implicit multifiel multiphase Eulerien recriticality (SIMMER) III, and SIMMER V simulation tools as well as comparison with its available experimental data. This program was performed in the 1980s in the frame of the Safety Assessment studies of Superphenix sodium-cooled reactor. This test was dedicated to verify the stability of a pure molten UO2 pool under decay heat conditions within natural convection and the long-term resilience of the peripheral fuel crust. The pool was generated in a stainless steel crucible by a progressive heating of a fuel pellet stack through six successive power plateaus. For the benchmark purposes, only the molten pool behavior k the last power plateau, where the pool was the largest and the fuel temperatures the highest, was investigated. Experimental data such as the axial profile of radial heat fluxes and heat transfer from the pool to the surrounding interassembly coolant as well as the axial profile of the peripheral fuel crust thickness were used for the reassessment of the simulation tools. In addition, other variables of interest not measured during the test, such as the radial and axial velocities in the pool, were also benchmarked. Finally, a critical analysis of the correlations and models used in the different simulation tools for the BF1 test modeling is also provided in the paper