Special Section Papers

Thermomechanical Behavior of Coolant Channel Assembly in Heavy Water Reactor Under Severe Plant Condition

[+] Author and Article Information
A. K. Dureja

Homi Bhabha National Institute,
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

P. Seshu

Department of Mechanical Engineering,
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

R. K. Sinha

Homi Bhabha Chair Professor
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 3(2), 020905 (Mar 01, 2017) (8 pages) Paper No: NERS-16-1093; doi: 10.1115/1.4035784 History: Received August 26, 2016; Revised December 17, 2016

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.

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Fig. 1

Primary heat transport system for pressurized heavy water reactor [1]

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Fig. 2

Coolant channel assembly for pressurized heavy water reactor [2]

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Fig. 3

Temperature dependency of YS of Zr-2.5Nb alloy PT material

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Fig. 4

Temperature dependency of UTS of Zr-2.5Nb alloy PT material

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Fig. 5

Flow curves for Zr-2.5Nb alloy PT material up to operating temperature

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Fig. 6

Flow curves for Zr-2.5Nb alloy PT material at higher temperatures

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Fig. 7

Comparison of fresh test specimen and specimen tested at 750 °C

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Fig. 8

(a) Schematic arrangement of axial heat flow apparatus and (b) temperature profile in axial heat flow apparatus

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Fig. 9

Experimental facility to measure thermal contact conductance

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Fig. 10

Actual variation and linear fit of TCC data for Zr-2.5Nb alloy PT sample with contact pressure

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Fig. 11

Actual variation and linear fit of TCC for Zr-4 CT sample with contact pressure

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Fig. 12

Variation of TCC with contact pressure for PT–CT sample

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Fig. 13

General arrangement of PT–CT assembly in the test setup [14]

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Fig. 14

Two-dimensional plane strain model for prediction of contact time between PT and CT

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Fig. 15

True stress–plastic strain data for PT material (Zr-2.5Nb alloy) as a function of temperature [10]

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Fig. 16

Radial deformation of pressure tube with time for 2 MPa internal pressure

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Fig. 17

Radial deformation of pressure tube with time for 4 MPa internal pressure

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Fig. 18

Time–temperature curve for contacting calandria tube for extreme cases of thermal contact conductance



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