Abstract
The first stage of a core degradation—based on the defense-in-depth concept of nuclear power plant (NPP) safety—is prone to fuel melting due to local blockage. The flow blockage accidents with no SCRAM happening can lead to a local fuel-clad failure, consequently affecting the safety of NPP. The present study provides an analysis of Anticipated Transient Without SCRAM (ATWS), which might lead to a condition of burning out. The accidents related to the ATWS scenarios, detailed in the case of VVER-1000/V446 reactor FSAR (Final Safety Analysis Report), include pump failure, local blockage, relative power increase, and a combination of these transients. In this research, first, drawing upon MCNPX 2.7 and COBRA-EN codes, a coupling framework is developed and then validated using an authentic reference point. The obtained results reveal that the reactor SCRAM does not occur while accidents are being investigated as there is a 10% difference in the mass flow rate reduction, a 470 kPa in the channel pressure drop, and a 204°K in the clad temperature, which constitute limitations under most pessimistic scenarios. However, under these conditions, a 70% void fraction over 12 min is observed in certain channels. Hence, burnout and local fuel melting could occur under normal operational and ATWS circumstances. According to uncertainty analyses, the occurrence of the void fraction above zero is locally definite. The transient analysis outputs could be deployed as monitoring system inputs and exploited for identifying weak points in the system.
Highlights
The ratio of heat removal/heat generation to the local or generalized flow reduction is the most significant and hazardous condition that threatens the safety of a nuclear power plant (NPP)
We provide the research results on the core relative power, fuel, clad, coolant temperatures, core pressure, void fractions, and MDNBR quantities during Anticipated Transient Without SCRAM (ATWS) scenarios implemented in the COBRA-EN code and MCNPX 2.7
The present study was designed to carry out transient analyses of scenario accidents that might lead to partial or local melting of the fuel clad while the reactor was in operation, that is, a situation far from SCRAM condition
Summary
The ratio of heat removal/heat generation to the local or generalized flow reduction is the most significant and hazardous condition that threatens the safety of a nuclear power plant (NPP). This could happen locally due to a flow blockage accident, including pump failure, or blockage at the channel entrance or in the middle of cooling channels between fuel rods. The local flow blockage accidents have a local characteristic and do not bring about any changes in total reactivity, core transient flow, or total heat absorption.
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