Investigating of Strain Effect on Cleavage Fracture for Reactor Pressure Vessel Material

Failure mechanism of 20MnMoNi55 steel in the lower self of ductile to brittle transition (DBT) region is considered as brittle fracture but it has been observed from the experimental analysis of stress-strain diagram that clear plastic deformation is shown by the material before failure. Therefore, strain correction is implemented in the cleavage fracture model proposed by different researchers in the lower self of the DBT region with the help of finite element analysis. To avoid a huge number of experiments being performed, Monte Carlo simulation is used to generate a huge number of random data at different temperatures in the lower self of the DBT region for calibration of the cleavage parameters with the help of the master curve methodology. Fracture toughness calculated after strain correction through different models are validated with experimental results for the different probability of failures.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.