Resulting from a mechanical failure of a control rod system the reactor coolant pressure ejects the control rod assembly from the core. The control rod ejection accident belongs to the category of reactivity initiated accident (RIA) and is defined as a design basis accident. The U.S. Nuclear Regulatory Commission (U.S. NRC) has established the Phenomenon Identification and Ranking Tables (PIRT) for rod ejection accident in pressurized water reactor (PWR) which will be used in this study. The present work is focused on selected physical parameters which can be calculated by coupling two best estimate codes for VVER-1000/V-320. Namely, TRACE V5 p4 was used to develop a thermal-hydraulics model and PARCS 3.2 for establishing a detailed neutronics model of a reactor core. Both models were executed in a merged version TRACE V5 p4. The analysis was focused on evaluation of the following parameters according to PIRT: relative power in the reactor core and rejected fuel assembly, inserted reactivity, moderator properties (density and temperature), and inner/outer temperature of the heat structure. Nominal power of 100% and the control rod ejection time of 0.1 s were considered. The inserted reactivity has reached 0.45$ and the maximum power in the reactor core has risen to 179.7%.
- RIA accident
ASJC Scopus subject areas
- Nuclear Energy and Engineering
- Safety, Risk, Reliability and Quality
- Energy Engineering and Power Technology
- Waste Management and Disposal