Nuclear Engineering and Design 255 (2013) 138–145 Contents lists available at SciVerse ScienceDirect Nuclear Engineering and Design j ourna l ho me pag e: www.elsevier.com/locate/nucengdes Experimental simulation of asymmetric heat up of coolant channel under small break LOCA condition for PHWR Ashwini K. Yadav a,1 , P. Majumdar b,2 , Ravi Kumar a,∗ , B. Chatterjee b,3 , Akhilesh Gupta a,4 , D. Mukhopadhyay b,5 a Department of Mechanical & Industrial Engineering, Indian Institute of Technology, Roorkee 247667, India b Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085, India h i g h l i g h t s I Circumferential temperature gradient of PT for asymmetric heat-up was 440 ◦ C. I At 2 MPa ballooning initiated at 450 ◦ C and with strain rate of 0.0277%/s. I At 4 MPa ballooning initiated at 390 ◦ C and with strain rate of 0.0305%/s. I At 4 MPa, PT ruptured under uneven strain and steep temperature gradient. I Integrity of PT depends on internal pressure and magnitude of decay power. a r t i c l e i n f o Article history: Received 29 May 2012 Received in revised form 31 October 2012 Accepted 1 November 2012 a b s t r a c t 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 Abbreviations: CANDU, Canadian Deuterium Uranium; LOCA, loss of coolent accident; ECCS, Emergency Core Cooling System; IPHWR, Indian Pressurised Heavy Water Reactor; PT, pressure tube; CT, Calandria Tube. ∗ Corresponding author. Tel.: +91 1332 285740/285117; fax: +91 1332 285665/273560. E-mail addresses: ashwinikumaryadav@gmail.com (A.K. Yadav), pmajum@barc.gov.in (P. Majumdar), ravikfme@iitr.ernet.in (R. Kumar), barun@barc.gov.in (B. Chatterjee), akhilfme@iitr.ernet.in (A. Gupta), dmukho@barc.gov.in (D. Mukhopadhyay). 1 Tel.: +91 8791203477. 2 Tel.: +91 22 25595174. 3 Tel.: +91 22 25595184. 4 Tel.: +91 1332 285613. 5 Tel.: +91 22 25593776. 0029-5493/$ – see front matter © 2012 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.nucengdes.2012.11.002