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Calandria Tube
The Experts below are selected from a list of 231 Experts worldwide ranked by ideXlab platform
R.k. Sinha – 1st expert on this subject based on the ideXlab platform
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Analysis of Interaction of Calandria Tubes with the End Shields in Advanced Heavy Water Reactor (AHWR)
Procedia Engineering, 2020Co-Authors: A.k. Dureja, Sucharita Sinha, R.k. SinhaAbstract:Abstract Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure Tube type, boiling light water cooled and heavy water moderated reactor. It consists of 452 coolant channels wherein the heat of nuclear reaction is used to boil the light water coolant to generate steam. Typical coolant channel assembly of AHWR consists of a pressure Tube of cold worked Zr- 2.5Nb alloy extended by stainless steel end-fittings, attached at its ends. A Calandria Tube of annealed Zircaloy-4 alloy separates the hot pressure Tube from the cold heavy water moderator. The Calandria Tube is attached at its ends to the Calandria side Tube-sheets of both the bottom and top end shields. Thermal expansion, irradiation enhanced axial creep and growth of Calandria Tubes cause them to elongate. These elongations induce compressive stresses in Calandria Tubes and also load the Tube-sheets of the end shields. Stresses also get induced in the annular plate, main-shell and sub-shells of the Calandria vessel. Nature of interaction and types of stresses induced in these components will depend upon the support conditions of both the end shields, creep and growth rates of Calandria Tubes. This has been studied for different combinations of fixed and floating types of support conditions of both the top and the bottom end shields. It has been observed that compared to the fixed support condition of both the end shields, stresses induced in the different components reduce to half when either of the end shields support condition is kept floating or on some flexible support.
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Thermo-mechanical Behaviour of Coolant Channels for Heavy Water Reactors under Accident Conditions
, 2020Co-Authors: Barc Newsletter, A.k. Dureja, P. Seshu, R.k. SinhaAbstract:The objective of nuclear safety research programme is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off normal conditions. Indian Pressurised 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 thermo-mechanical behaviour. 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. Since the emergency core cooling system 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/balloon 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 inter-dependency between thermal and mechanical contact behaviour. The amount of heat thus expelled significantly depends on the thermal contact conductance 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 thermo-mechanical 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 has been modeled using finite element based program. Experimental data of pressure Tube material, generated for this research work, were used in modelling and examining the role of nonlinear stress-strain laws in the finite element analyses.
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Thermo-mechanical Behaviour of Coolant Channel Assembly in Heavy Water Reactor under Severe Plant Condition
Journal of Nuclear Engineering and Radiation Science, 2017Co-Authors: Adarsh K. Dureja, D. N. Pawaskar, Seshu Pasumarthy, R.k. SinhaAbstract: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.
Thomas W. Krause – 2nd expert on this subject based on the ideXlab platform
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Evaluation of Concentric Tube Model for Pressure Tube to Calandria Tube Gap Measurement
IEEE Sensors Journal, 2019Co-Authors: Geoffrey Klein, Jordan E. Morelli, Mark S. Luloff, Thomas W. KrauseAbstract:The CANDU® nuclear reactor fuel channels consist of a pressure Tube (PT) contained within a Calandria Tube (CT). The gap between the hot (~300° C) PT and cooler (~50° C) CT is required since the contact could lead to hydride blister formation with subsequent cracking. This gap is measured using a drive-receive eddy current technology, providing assurance that the contact is not imminent. The analytical models of probe response to gap use flat plates to approximate the PT and CT geometry. In this paper, a semi-analytical model that approximates the PT within the CT geometry as two concentric Tubes with correct PT curvature, but a CT diameter, which is adjusted to change gap, is compared with a flat-plate model of probe response to changing gap against experimental measurements using standard CT and PT samples, with varying wall thickness and resistivity that simulates in-reactor variation of these parameters. The concentric and flat plate models both predict gap values with an average error of 0.1 mm between contact and 5 mm gap at 4.2 kHz and 8.0 kHz. The concentric model has an average relative gap error at the maximum gap of 2% at 4.2 kHz and 8.0 kHz, whereas that of the flat-plate model is 10% and 12%, respectively. The improved accuracy of the concentric model at larger gaps is attributed to the incorporation of Tube curvature, which limits the probe’s magnetic field spread when compared to the flat-plate model.
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Principal Components Analysis of Multifrequency Eddy Current Data Used to Measure Pressure Tube to Calandria Tube Gap
IEEE Sensors Journal, 2016Co-Authors: Shaddy Shokralla, Jordan E. Morelli, Thomas W. KrauseAbstract:Principal components analysis (PCA) involves transforming a set of correlated observations into a set of linearly uncorrelated variables, which can reveal simplified trends in data. Multifrequency eddy current testing contains correlations across different test frequencies. In this paper, PCA is used to extract unique information from multifrequency eddy current data sets, used to measure the pressure Tube to Calandria Tube gap, in CANada Deuterium Uranium fuel channels. Advantages include compressed data acquisition, allowing for increased inspection speed, and monitoring for variation in physical parameters using a reduced number of variables. PCA employing analytical input model data is validated against PCA employing data from physical experiments.
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modelling and validation of eddy current response to changes in factors affecting pressure Tube to Calandria Tube gap measurement
Ndt & E International, 2015Co-Authors: Shaddy Shokralla, Sean Sullivan, Jordan Morelli, Thomas W. KrauseAbstract:Abstract Procedures employed to non-destructively examine nuclear power plants must undergo inspection qualification to ensure that they meet their respective inspection specification requirements. Modelling is a powerful tool that can be exploited in the inspection qualification process. The gap between pressure Tubes (PTs) and Calandria Tubes (CTs) in CANDU (CANada Deuterium Uranium) fuel channels is periodically measured, as contact can result in localized cooling and potential cracking. This work shows how an analytical model can be employed to characterize the effects of PT wall thickness and resistivity variation on gap measurement, and details its validation against physical experiments.
Shaddy Shokralla – 3rd expert on this subject based on the ideXlab platform
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comprehensive characterization of measurement data gathered by the pressure Tube to Calandria Tube gap probe
, 2016Co-Authors: Shaddy ShokrallaAbstract:Thesis (Ph.D, Physics, Engineering Physics and Astronomy) — Queen’s University, 2016-08-29 20:13:30.327
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Principal Components Analysis of Multifrequency Eddy Current Data Used to Measure Pressure Tube to Calandria Tube Gap
IEEE Sensors Journal, 2016Co-Authors: Shaddy Shokralla, Jordan E. Morelli, Thomas W. KrauseAbstract:Principal components analysis (PCA) involves transforming a set of correlated observations into a set of linearly uncorrelated variables, which can reveal simplified trends in data. Multifrequency eddy current testing contains correlations across different test frequencies. In this paper, PCA is used to extract unique information from multifrequency eddy current data sets, used to measure the pressure Tube to Calandria Tube gap, in CANada Deuterium Uranium fuel channels. Advantages include compressed data acquisition, allowing for increased inspection speed, and monitoring for variation in physical parameters using a reduced number of variables. PCA employing analytical input model data is validated against PCA employing data from physical experiments.
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modelling and validation of eddy current response to changes in factors affecting pressure Tube to Calandria Tube gap measurement
Ndt & E International, 2015Co-Authors: Shaddy Shokralla, Sean Sullivan, Jordan Morelli, Thomas W. KrauseAbstract:Abstract Procedures employed to non-destructively examine nuclear power plants must undergo inspection qualification to ensure that they meet their respective inspection specification requirements. Modelling is a powerful tool that can be exploited in the inspection qualification process. The gap between pressure Tubes (PTs) and Calandria Tubes (CTs) in CANDU (CANada Deuterium Uranium) fuel channels is periodically measured, as contact can result in localized cooling and potential cracking. This work shows how an analytical model can be employed to characterize the effects of PT wall thickness and resistivity variation on gap measurement, and details its validation against physical experiments.