Antoaneta Stefanova

Investigation of Quench-16 Experiment With MELCOR Computer Code

This paper presents a comparison of MELCOR calculated results with experimental data for the QUENCH-16 experiment. The analysis for the air ingress experiment QUENCH-16 has been performed by INRNE. The calculations have been performed with MELCOR code. The QUENCH-16 experiment has been performed on 27-th of July 2011 in the frame of the EC-supported LACOMECO program. The experiments have focused on air ingress investigation into an overheated core following earlier partial oxidation in steam. QUENCH-16 has been performed with limited pre-oxidation and low air flow rate. One of the main objectives of QUENCH-16 was to examine the interaction between nitrogen and oxidized cladding during a prolonged period of oxygen starvation. The bundle is made from 20 heated fuel rod simulators arranged in two concentric rings and one unheated central fuel rod simulator, each about 2.5 m long. The tungsten heaters were surrounded by annular ZrO2 pellets to simulate the UO2 fuel. The geometry and most other bundle components are prototypical for Western-type PWRs. To improve the obtained results it has been made a series of calculations to select an appropriate initial temperature of the oxidation of the fuel bundle and modified correlation oxidation of Zircaloy with MELCOR computer code. The compared results have shown good agreement of calculated hydrogen and oxygen starvation in comparison with test data.Copyright

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel.In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed.The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident.The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research.The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.Copyright

Simulation of Quench-12 Test with ASTEC1.3.2 Computer Code

This paper presents an application of the ASTEC V1.3R2 computer code for simulation of QUENCH-12 experiment. The test have been performed to investigate the behavior of VVER type of fuel assemblies during severe accident conditions. In the performed analyses it have been assessed the mass of generated hydrogen during the experiment flooding of overheated core. The base line input model for ASTEC has been provided by Forschungszentrum, Karlsruhe. The comparison of ASTEC1.3R2 calculated results with measured test data shows good agreement.