Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR

[+] Author and Article Information
D. Mukhopadhyay, P. Majumdar, G. Behera, S. K. Gupta, V. Venkat Raj

Health, Safety and Environment Group, Bhabha Atomic Research Centre, Reactor Safety Division, BARC, Mumbai 400 085, India

J. Pressure Vessel Technol 124(2), 161-167 (May 01, 2002) (7 pages) doi:10.1115/1.1463036 History: Received December 04, 2000; Revised December 18, 2001; Online May 01, 2002
Copyright © 2002 by ASME
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Fischer, S. R., 1978, “RELAP4/MOD6: A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems, User’s Manual,” Technical Report No. CDAP TR003, Idaho National Engineering Laboratory (INEL), ID.
Gupta, S. K., Venkat Raj, V., and Kakodkar, A., 1996, “A Study of Indian PHWR Reactor Channel Under Prolonged Deteriorated Flow Conditions,” Proc., IAEA TCM on Advances in Heavy Water Reactors, Bombay, India.
Shewfelt,  R. S. W., Lyall,  L. W., and Godin,  D. P., 1984, “A High Temperature Creep Model for Zr-2.5 wt percent Nb Pressure Tubes,” J. Nucl. Mater., 125, pp. 228–235.


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Presure transients of the feeders
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Mass flow rate transients
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Clad temperature of fuel pin no. 2 at channel center
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Clad temperature of fuel pin no. 5 at channel center
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Clad temperature of fuel pin no. 8 at channel center
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Clad temperature of fuel pin no. 14 at channel center
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Temperature of PT and CT at channel center
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Schematic of primary heat transport (PHT) system
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Cross section of a reactor channel of 220 MWe PHWR
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Nodalization scheme used for PHT system simulation of PHWR with RELAP4/MOD6
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Different types of subchannels used in HT/MOD4 Code
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Equivalent network for n segmented enclosure
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Nodalization for the 19 fuel pins along with PT and CT showing fuel pin numbers and circumferential nodes
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Clad temperature transients for different break sizes
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Header pressure transients
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Circumferencial creep strain of the PT at channel center
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Circumferential temperature distribution of PT at channel center



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