• Sandip Ghosh

The leak flow estimation through a narrow channel is important for ascertaining safe designs of piping system in nuclear power plants. RELAP is a popular code for predicting the thermal-hydraulic behavior of leakage flow during loss of coolant accidents. In this paper leak flow through narrow slits created on a reactor grade pipe has been simulated and code predictions have been analyzed. Pressurized subcooled light water was used to simulate standard conditions of pressurized water reactors with a maximum pressure of 90 Bar. For various stagnation conditions and pipe geometries, experimental results from literature have been compared with code predictions. Leak Flow behavior has been studied for varying slit opening geometries and maximum temperature conditions.

Narrow slit, LWR, choked flow, RELAP5, LOCA.

"Thermal-hydraulic Code prediction of primary loop leakage in NPP ", International Journal of Emerging Technologies and Innovative Research (www.jetir.org), ISSN:2349-5162, Vol.6, Issue 5, page no.162-167, May-2019, Available :http://www.jetir.org/papers/JETIR1905027.pdf

"Thermal-hydraulic Code prediction of primary loop leakage in NPP ", International Journal of Emerging Technologies and Innovative Research (www.jetir.org | UGC and issn Approved), ISSN:2349-5162, Vol.6, Issue 5, page no. pp162-167, May-2019, Available at : http://www.jetir.org/papers/JETIR1905027.pdf