The objective of current study is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off-normal conditions. Indian pressurized heavy water reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermomechanical behavior. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors, i.e., PHWRs, is loss of coolant accident (LOCA) coincident with loss of emergency core cooling system (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low- or no-flow condition and inventory depletion of primary side. Initially, this will result in high temperature of the fuel pins. Since the emergency core cooling system (ECCS) is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment, and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube—calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag (by weight of fuel bundle) and/or balloon (by internal pressure) and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing interdependency between thermal and mechanical contact behavior. The amount of heat thus expelled significantly depends on the thermal contact conductance (TCC) and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively cold moderator. This, in turn, limits the temperature of fuel for a sufficiently long period and ensures safety of the plant. The objective of this paper is to provide insights into this thermomechanical behavior by computational studies and to understand the role of underlying parameters (such as material constants, thermal contact conductance, and boundary conditions) that control the tube deformation and further damage progression. The deformation characteristics of the pressure tube have been modeled using finite-element-based program. Experimental data of pressure tube material, generated for this research work, were used in modeling and examining the role of nonlinear stress–strain laws in the finite-element analyses.
Skip Nav Destination
Article navigation
April 2017
Research-Article
Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition
A. K. Dureja,
A. K. Dureja
Homi Bhabha National Institute,
R. No. 209, Training School Complex,
Anushaktinagar, Mumbai 94, India;
R. No. 209, Training School Complex,
Anushaktinagar, Mumbai 94, India;
Bhabha Atomic Research Centre,
Trombay, Mumbai 85, India
e-mail: dureja@hbni.ac.in; akdureja@barc.gov.in
Trombay, Mumbai 85, India
e-mail: dureja@hbni.ac.in; akdureja@barc.gov.in
Search for other works by this author on:
P. Seshu,
P. Seshu
Department of Mechanical Engineering,
Indian Institute of Technology Bombay,
IITB, Powai, Mumbai 76, India
e-mail: seshu@iitb.ac.in
Indian Institute of Technology Bombay,
IITB, Powai, Mumbai 76, India
e-mail: seshu@iitb.ac.in
Search for other works by this author on:
D. N. Pawaskar,
D. N. Pawaskar
Department of Mechanical Engineering,
Indian Institute of Technology Bombay,
IITB, Powai,
Mumbai 76, India
e-mail: pawaskar@iitb.ac.in
Indian Institute of Technology Bombay,
IITB, Powai,
Mumbai 76, India
e-mail: pawaskar@iitb.ac.in
Search for other works by this author on:
R. K. Sinha
R. K. Sinha
Homi Bhabha Chair Professor
Department of Atomic Energy,
BARC, Central Complex,
Trombay, Mumbai 85, India
e-mail: rksinha@barc.gov.in
Department of Atomic Energy,
BARC, Central Complex,
Trombay, Mumbai 85, India
e-mail: rksinha@barc.gov.in
Search for other works by this author on:
A. K. Dureja
Homi Bhabha National Institute,
R. No. 209, Training School Complex,
Anushaktinagar, Mumbai 94, India;
R. No. 209, Training School Complex,
Anushaktinagar, Mumbai 94, India;
Bhabha Atomic Research Centre,
Trombay, Mumbai 85, India
e-mail: dureja@hbni.ac.in; akdureja@barc.gov.in
Trombay, Mumbai 85, India
e-mail: dureja@hbni.ac.in; akdureja@barc.gov.in
P. Seshu
Department of Mechanical Engineering,
Indian Institute of Technology Bombay,
IITB, Powai, Mumbai 76, India
e-mail: seshu@iitb.ac.in
Indian Institute of Technology Bombay,
IITB, Powai, Mumbai 76, India
e-mail: seshu@iitb.ac.in
D. N. Pawaskar
Department of Mechanical Engineering,
Indian Institute of Technology Bombay,
IITB, Powai,
Mumbai 76, India
e-mail: pawaskar@iitb.ac.in
Indian Institute of Technology Bombay,
IITB, Powai,
Mumbai 76, India
e-mail: pawaskar@iitb.ac.in
R. K. Sinha
Homi Bhabha Chair Professor
Department of Atomic Energy,
BARC, Central Complex,
Trombay, Mumbai 85, India
e-mail: rksinha@barc.gov.in
Department of Atomic Energy,
BARC, Central Complex,
Trombay, Mumbai 85, India
e-mail: rksinha@barc.gov.in
Manuscript received August 26, 2016; final manuscript received December 17, 2016; published online March 1, 2017. Assoc. Editor: Arun Nayak.
ASME J of Nuclear Rad Sci. Apr 2017, 3(2): 020905 (8 pages)
Published Online: March 1, 2017
Article history
Received:
August 26, 2016
Revised:
December 17, 2016
Citation
Dureja, A. K., Seshu, P., Pawaskar, D. N., and Sinha, R. K. (March 1, 2017). "Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition." ASME. ASME J of Nuclear Rad Sci. April 2017; 3(2): 020905. https://doi.org/10.1115/1.4035784
Download citation file:
Get Email Alerts
Cited By
Operation Optimization Framework for Advanced Reactors Using a Data-Driven Digital Twin
ASME J of Nuclear Rad Sci (April 2025)
Numerical Analysis of Gas Generation and Migration in a Radioactive Waste Disposal Cell of a Deep Geological Repository
ASME J of Nuclear Rad Sci (April 2025)
Related Articles
Thermomechanical Behavior of Pressure Tube Under Small Break Loss of Coolant Accident for PHWR
J. Pressure Vessel Technol (August,2013)
Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop
ASME J of Nuclear Rad Sci (April,2018)
Gap Formation and Interfacial Heat Transfer Between Thermoelastic Bodies in Imperfect Contact
J. Heat Transfer (April,2001)
Large Strain Mechanical Behavior of HSLA-100 Steel Over a Wide Range of Strain Rates
J. Eng. Mater. Technol (January,2012)
Related Proceedings Papers
Related Chapters
Microstructure Evolution and Physics-Based Modeling
Ultrasonic Welding of Lithium-Ion Batteries
Compressive Deformation of Hot-Applied Rubberized Asphalt Waterproofing
Roofing Research and Standards Development: 10th Volume
Part 2, Section II—Materials and Specifications
Companion Guide to the ASME Boiler & Pressure Vessel Code, Volume 1, Second Edition