Abstract
In a nuclear reactor temperature rises drastically in fuel channels under loss of coolant accident due to failure of primary heat transportation system. Present investigation has been carried out to capture circumferential and axial temperature gradients during fully and partially 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 circumferential temperature gradient over PT and could lead to breaching of PT under high pressure.
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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
References
Brown RA, Blahni C, Mujumdar AP (1984) Degraded cooling in a CANDU reactor. Nucl Sci Eng 88(3):425–435
Gillespie GE, Moyer RG, Litke D C (1987) The experimental determination of circumferential temperature distributions developed in pressure tube during slow coolant boil down. In: Proc. CNS 8th Annual Conference, Saint John, pp 241–248
Gulshani P (1987) Prediction of pressure tube integrity for a small LOCA and total loss of emergency coolant injection in CANDU. Trans Am Nucl Soc 55:461
Gupta SK, Dutta BK, Venkatraj V, Kakodkar A (1996) A study of Indian PHWR reactor channel under prolonged deteriorated flow conditions. In: IAEA TCM on advances in heavy water reactor. India, Bhabha Atomic Research Centre
Hanna BN (1998) CATHENA: a thermalhydraulic code for CANDU analysis. Nucl Eng Des 180:113–131
Kohn E, Hadaller GI, Sawala RM, Archinoff GH, Wadsworth SL (1985) CANDU fuel development during severely degraded cooling: experimental results. In: Canadian Nuclear Society Conference, Ottawa, Ontario
Kuehn TH, Goldstein RJ (1976) An experimental and theoretical study of natural convection in the horizontal annulus between horizontal concentric cylinders. J Fluid Mech 74(4):695–719
Majumdar P, Mukhopadhyay D, Gupta SK, Kushwaha HS, Venkat Raj V (2004) Simulation of pressure tube deformation during high temperature transients. Int J Press Vessels Pip 81(7):575–581
Nandan G, Sahoo PK, Kumar R, Chatterjee B, Mukhopadhyay D, Lele HG (2010) Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition. Nucl Eng Des 240(10):3504–3512
Nandan G, Sahoo PK, Kumar R, Chatterjee B, Mukhopadhyay D, Lele HG (2010) Thermo-mechanical behavior of pressure tube of Indian PHWR at 20 Bar pressure. World Academy of Science, Engineering and Technology 37
NandanG MP, Sahoo PK, Kumar R, Chatterjee B, Mukhopadhyay D, Lele HG (2012) Study of ballooning of a completely voided pressure tube of Indian PHWR under heat-up condition. Nucl Eng Des 243:3504–3512
Rodchenkov BS, Semenov AN (2005) High temperature mechanical behaviour of Zr-2.5% Nb alloy. Nucl Eng Des 235:2009–2018
Shewfelt RSW, Layall LW, Godin DP (1984) High temperature creep model for Zr-2.5 wt% Nb pressure tubes. J Nucl Mater 125:228–235
Shewfelt RSW, Lyall LW (1985) A high temperature longitudinal strain rate equation for Zr–2.5 wt% Nb pressure tubes. J Nucl Mater 132:41–46
Yuen PS, So CB, Moyer RG, Litke DC (1988) The experimental measurement of circumferential temperature distributions developed on pressure tubes under stratified two-phase of conditions. In Proc. CNS 9th Annual Conference, Winnipeg, Manitoba, pp 120–126
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The authors are thankful to Bhabha Atomic Research Centre (BARC), Mumbai, India for providing financial support to this project.
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Yadav, A.K., kumar, R., Gupta, A. et al. Experimental investigation on circumferential and axial temperature gradient over fuel channel under LOCA. Heat Mass Transfer 50, 737–746 (2014). https://doi.org/10.1007/s00231-013-1279-8
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DOI: https://doi.org/10.1007/s00231-013-1279-8