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Please use this identifier to cite or link to this item: http://repository.iitr.ac.in/handle/123456789/11532
Title: Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR
Authors: Yadav A.K.
Majumdar P.
Kumar R.
Chatterjee B.
Gupta, Akhilesh Kumar
Mukhopadhyay D.
Published in: Nuclear Engineering and Design
Abstract: During postulated small break loss of coolant accident (SBLOCA) for Pressurised Heavy Water Reactors (PHWRs) as well as for postulated SBLOCA coincident with loss of ECCS, a stratified flow condition can arise in the coolant channels as the gravitational force dominates over the low inertial flow arising from small break flow. A Station Blackout condition without operator intervention can also lead to stratified flow condition during a slow channel boil-off condition. For all these conditions the pressure remains high and under stratified flow condition, the horizontal fuel bundles experience different heat transfer environments with respect to the stratified flow level. This causes the bundle upper portion to get heated up higher as compared to the submerged portion. This kind of asymmetrical heating of the bundle is having a direct bearing on the circumferential temperature gradient of pressure tube (PT) component of the coolant channel. The integrity of the PT is important under normal conditions as well as at different accident loading conditions as this component houses the fuel bundles and serves as a coolant pressure boundary of the reactors. An assessment of PT is required with respect to different accident loading conditions. The present investigation aims to study thermo-mechanical behaviour of PT (Zr, 2.5 wt% Nb) under a stratified flow condition under different internal pressures. The component is subjected to an asymmetrical heat-up conditions as expected during the said situation under different pressure conditions which varies from 2.0 MPa and 4 MPa. In order to simulate partially voided conditions inside PT, asymmetric heating has been carried out by injecting power to selected heater pins of the upper section of the 19 element fuel bundle simulator housed in a PT. This simulates nearly a stratification level of a half filled reactor channel. Through this technique an expected maximum circumferential temperature gradient of around 440 °C, has been attended from top to bottom periphery of PT. Tests also cover a power range of 8-11 kW which simulates different decay power levels. An asymmetric ballooning over eighty percent of PT length is observed for all the experiments and the deformation is mostly located to the upper part of the PT. The PT integrity is observed for lower internal pressure tests however a local failure has been observed for the test at 4.0 MPa. This is found to be due to excessive local strain prior to establishment of contact with Calandria Tube. © 2012 Elsevier B.V.
Citation: Nuclear Engineering and Design (2013), 255(): 138-145
URI: https://doi.org/10.1016/j.nucengdes.2012.11.002
http://repository.iitr.ac.in/handle/123456789/11532
Issue Date: 2013
ISSN: 295493
Author Scopus IDs: 55520507000
35613969100
55389796000
7201648525
55491955100
7102685834
Author Affiliations: Yadav, A.K., Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667, India
Majumdar, P., Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085, India
Kumar, R., Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667, India
Chatterjee, B., Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085, India
Gupta, A., Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667, India
Mukhopadhyay, D., Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085, India
Corresponding Author: Kumar, R.; Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee 247667, India; email: ravikfme@iitr.ernet.in
Appears in Collections:Journal Publications [ME]

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