INVESTIGATION OF THERMAL HYDRAULICS PARAMETERS OF VVER-1000/V-320 DURING LOCA ACCIDENT | ||||
| Arab Journal of Nuclear Sciences and Applications | ||||
| Volume 58, Issue 1, January 2025, Page 65-73 PDF (717.83 K) | ||||
| Document Type: Original Article | ||||
| DOI: 10.21608/ajnsa.2024.321389.1853 | ||||
 View on SCiNiTO | ||||
| Authors | ||||
Karim GadEl-karim1; Alya Badawi2; Ahmed Emad El-Din El-Saghir3; Asmaa Gamal Abo Elnour   4
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| 1Nuclear and radiation Engineering, Alexanderia Engineering Faculty, Alexandria, Egypt | ||||
| 2Nuclear & Radiation Engineering Department, Alexandria University, Alexandria, Egypt | ||||
| 3Nucl. Eng. Dept., Faculty of Eng. Alexandria University, Alexandria, Egypt | ||||
| 4Atomic Energy Authority Egypt | ||||
| Abstract | ||||
| This paper presents a model of the VVER-1000 reactor using data from KNPP/ VVER -1000/ V-320, the VVER-1000 coolant transient benchmark, and previous studies. The RELAP5/MOD3.2 code, commonly used in thermal hydraulics simulations, was utilized. Model validation included a steady-state analysis, with results closely aligning with the plant specified parameters. Following validation, LOCA scenario was simulated to evaluate its impact on thermal hydraulic parameters and the reactor’s behavior during severe accident conditions. In the LOCA simulation, the core region was identified as the most critical area for detecting thermal behavior during severe accident conditions. The cladding temperature distribution indicated a rise in temperature within the first seconds of the accident. Comparing temperatures between normal operation and LOCA conditions highlighted the behavior of the primary coolant during an accident. The pressure distribution across the core region revealed a pressure drop in the primary circuit and core dewatering. Overall, the RELAP5 model results demonstrated a good approximation to the plant behavior during an accident. | ||||
| Keywords | ||||
| VVER-1000/V-320; Thermal Hydraulics; RELAP5-3D; MCP trip; LOCA | ||||
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