ORIGINAL Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA Ashwini Kumar Yadav • Ravi kumar • Akhilesh Gupta • Barun Chatterjee • Deb Mukhopadhyay • H. G. Lele Received: 5 December 2011 / Accepted: 15 December 2013 / Published online: 25 December 2013 Ó Springer-Verlag Berlin Heidelberg 2013 Abstract In a nuclear reactor temperature rises drasti- cally in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferen- tial and axial temperature gradients during fully and par- tially voiding conditions in a fuel channel using 19 pin fuel element simulator. A series of experiments were carried out by supplying power to outer, middle and center rods of 19 pin fuel simulator in ratio of 1.4:1.1:1. The temperature at upper periphery of pressure tube (PT) was slightly higher than at bottom due to increase in local equivalent thermal conductivity from top to bottom of PT. To simulate fully voided conditions PT was pressurized at 2.0 MPa pressure with 17.5 kW power injection. Ballooning initiated from center and then propagates towards the ends and hence axial temperature difference has been observed along the length of PT. For asymmetric heating, upper eight rods of fuel simulator were activated and temperature difference up-to 250 °C has been observed from top to bottom periphery of PT. Such situation creates steep circumfer- ential temperature gradient over PT and could lead to breaching of PT under high pressure. Abbreviations CANDU Canadian Deuterium Uranium LOCA Loss of coolant accident ECCS Emergency core cooling system IPHWR Indian pressurized heavy water reactor PT Pressure tube CT Calandria tube 1 Introduction A standardized 220 MW(e) Indian PHWR has 306 hori- zontal fuel channels in a figure-of-eight loop configura- tion,with one pair of steam generators and pumps in each bank of PHTS loop as shown in Fig. 1. In case of 220 MW(e) each fuel channel has 12 fuel bundles and each bundle consists of 19 fuel elements. The coolant flows through 306 horizontal channels which are housed in cal- andria vessel and submerged in heavy water called mod- erator. The coolant flows through half of the channels in one direction and in remaining channels in opposite direction. The carbon dioxide gas at atmospheric pressure is filled in gap between calandria tube and pressure tube for thermal insulation. The nuclear heat is removed from fuel bundles by heavy water coolant in primary circuit and transferred to secondary circuit in steam generators. The failure of pump discharge line, reactor inlet header, single feeder pipe failure etc. come under single failure event. LOCA along with failure of ECCS or LOCA along with failure of ECCS and moderator cooling system come under multiple failure events. For certain accident conditions A. K. Yadav (&) Á R. kumar Á A. Gupta Indian Institute of Technology Roorkee, Roorkee, India e-mail: ashwinikumaryadav@gmail.com R. kumar e-mail: ravikfme@iitr.ernet.in A. Gupta e-mail: akhilfme@iitr.ernet.in B. Chatterjee Á D. Mukhopadhyay Á H. G. Lele Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai, India e-mail: barun@barc.gov.in D. Mukhopadhyay e-mail: dmukho@barc.gov.in H. G. Lele e-mail: hglele@barc.gov.in 123 Heat Mass Transfer (2014) 50:737–746 DOI 10.1007/s00231-013-1279-8