Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition
Pressure tube (zirconium 2.5 wt.% Nb) serves as a pressure boundary for the coolant that
removes nuclear heat generated in the reactor core of Indian Pressurised Heavy Water …
removes nuclear heat generated in the reactor core of Indian Pressurised Heavy Water …
Study of ballooning of a completely voided pressure tube of Indian PHWR under heat up condition
In a nuclear reactor, loss of coolant accident (LOCA) considers wide range of postulated
damage or rupture of pipe in the heat transport piping system. In case of LOCA with/without …
damage or rupture of pipe in the heat transport piping system. In case of LOCA with/without …
CFD analysis of suspended debris during postulated severe core damage accident of PHWR
A steady state CFD analyses for suspended debris with two different configurations are
carried out for heat transfer assessment. Very low frequency postulated accident without …
carried out for heat transfer assessment. Very low frequency postulated accident without …
Influence of PT-CT contact on PHWR fuel channel thermal behaviour under accident condition–An experimental study
An experimental investigation has been performed to simulate a scenario of a beyond
design basis accident like LOCA along with the un-availability of ECCS. The investigation is …
design basis accident like LOCA along with the un-availability of ECCS. The investigation is …
Effect of eccentricity of pressure tube on circumferential temperature distribution of PHWR fuel bundle under postulated accident condition
For a large break Loss of Coolant Accident (LOCA) scenario with simultaneous failure of
Emergence Core Cooling System (ECCS), the heat removal capability of coolant decreases …
Emergence Core Cooling System (ECCS), the heat removal capability of coolant decreases …
Experimental and numerical study of phwr specific suspended debris
Abstract In 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR), under a postulated
scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident …
scenario of unmitigated Station Black Out (SBO) or a large break Loss of Coolant Accident …
Experimental investigation of radiation heat transfer in coolant channel under impaired cooling scenario for Indian PHWR
Postulated accidents like large break LOCA leads to expulsion of coolant in the primary heat
transport system thus voiding of reactor core initially. However, the reactor is shut down at …
transport system thus voiding of reactor core initially. However, the reactor is shut down at …
Thermal analysis of severe channel damage caused by a stagnation channel break in a PHWR
D Mukhopadhyay, P Majumdar… - J. Pressure …, 2002 - asmedigitalcollection.asme.org
The reactor channel of the horizontal core of pressurized heavy water reactors experiences
very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders …
very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders …
Full length channel pressure tube sagging study under postulated LOCA with un-availability of ECCS in an Indian PHWR
S Negi, R Kumar, P Majumdar… - Nuclear Engineering and …, 2017 - Elsevier
An experimental investigation consisting of two experiments was conducted to assess the
thermal behavior of the Indian-made full length Pressure Tube (PT) of an Indian PHWR …
thermal behavior of the Indian-made full length Pressure Tube (PT) of an Indian PHWR …
Investigation of the channel disassembly behaviour of Indian 200MWe PHWR–A numerical approach
In absence or non-functioning of a failsafe system, the postulated accident in Pressurised
heavy water nuclear reactor may lead to severe degradation of structural integrity of …
heavy water nuclear reactor may lead to severe degradation of structural integrity of …