Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition

G. Nandan, P.K. Sahoo, R. Kumar, B. Chatterjee, D. Mukhopadhyay, H.G. Lele

Research output: Contribution to journalArticle

15 Citations (Scopus)

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.
Original languageEnglish
Pages (from-to)3504-3512
Number of pages9
JournalNuclear Engineering and Design
Volume240
Issue number10
DOIs
Publication statusPublished - 2010

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tubes
Heavy water reactors
heavy water reactors
temperature
accident
accidents
Temperature
bundles
Loss of coolant accidents
Moderators
nuclear heat
Reactor cores
reactor safety
experiment
Experiments
loss of coolant
Heat sinks
Structural integrity
Cooling systems
reactor cores

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Nandan, G. ; Sahoo, P.K. ; Kumar, R. ; Chatterjee, B. ; Mukhopadhyay, D. ; Lele, H.G. / Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition. In: Nuclear Engineering and Design. 2010 ; Vol. 240, No. 10. pp. 3504-3512.
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Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition. / Nandan, G.; Sahoo, P.K.; Kumar, R.; Chatterjee, B.; Mukhopadhyay, D.; Lele, H.G.

In: Nuclear Engineering and Design, Vol. 240, No. 10, 2010, p. 3504-3512.

Research output: Contribution to journalArticle

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