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Please use this identifier to cite or link to this item: http://repository.iitr.ac.in/handle/123456789/18010
Title: Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition
Authors: Nandan G.
Sahoo, Pradeep K.
Kumar R.
Chatterjee B.
Mukhopadhyay D.
Lele H.G.
Published in: Nuclear Engineering and Design
Abstract: 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 Reactors (IPHWRs). Under postulated low frequency (<10-6 per year) accidents like Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS) injection, heatup of pressure tube (PT) combined with internal pressure and the weight of the fuel bundle may lead to deformation. The extent and nature of deformation is important from reactor safety point of view. An experimental set-up has been designed and fabricated to simulate sagging (downward deformation) of PT due to its own weight and the weight of fuel bundles for 220 MWe IPHWRs. Experiments are conducted at different heatup rates of voided PTs. It is observed that sagging initiates at a temperature around 450 °C. Contact between PT and calandria tube (CT) occurs at around 585-625 °C, respectively. Once PT-CT contact takes place, PT temperature either decreases or the temperature rise remains controlled whereas CT temperature keeps on increasing for next 20-30 s. The contact location in all the experiments was near the centre of the tube. Structural integrity of PT is retained (no breach) for all the experiments. The PT temperature rise is found to be arrested after the contact between PT and CT, thus establishing that moderator acts as an efficient heat sink for IPHWRs. © 2010 Elsevier B.V. All rights reserved.
Citation: Nuclear Engineering and Design, (2010), 3504- 3512
URI: https://doi.org/10.1016/j.nucengdes.2010.05.042
http://repository.iitr.ac.in/handle/123456789/18010
Issue Date: 2010
Keywords: Contact location
Emergency core cooling system
Experimental investigations
Experimental setup
Fuel bundle
Heat-up
Heat-up rates
Internal Pressure
Low frequency
Nuclear heat
Pressure boundary
Pressure tubes
Reactor safety
Temperature rise
Accidents
Coolants
Deformation
Experiments
Heavy water
Heavy water reactors
Pressure tube reactors
Tubes (components)
Zirconium
Loss of coolant accidents
ISSN: 295493
Author Scopus IDs: 57200563897
22835953900
55389796000
7201648525
7102685834
6602692797
Author Affiliations: Nandan, G., Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Roorkee, Uttarakhand 247667, India
Sahoo, P.K., Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Roorkee, Uttarakhand 247667, India
Kumar, R., Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Roorkee, Uttarakhand 247667, India
Chatterjee, B., Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai, India
Mukhopadhyay, D., Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai, India
Lele, H.G., Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai, India
Corresponding Author: Sahoo, P. K.; Department of Mechanical and Industrial Engineering, Indian Institute of Technology Roorkee, Roorkee, Uttarakhand 247667, India; email: sahoofme@iitr.ernet.in
Appears in Collections:Conference Publications [ME]

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