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R.k. Sinha - One of the best experts on this subject based on the ideXlab platform.

  • Analysis of Interaction of Calandria Tubes with the End Shields in Advanced Heavy Water Reactor (AHWR)
    Procedia Engineering, 2020
    Co-Authors: A.k. Dureja, Sucharita Sinha, R.k. Sinha
    Abstract:

    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.

  • Thermo-mechanical Behaviour of Coolant Channels for Heavy Water Reactors under Accident Conditions
    2020
    Co-Authors: Barc Newsletter, A.k. Dureja, P. Seshu, R.k. Sinha
    Abstract:

    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.

  • Thermo-mechanical Behaviour of Coolant Channel Assembly in Heavy Water Reactor under Severe Plant Condition
    Journal of Nuclear Engineering and Radiation Science, 2017
    Co-Authors: Adarsh K. Dureja, D. N. Pawaskar, Seshu Pasumarthy, R.k. Sinha
    Abstract:

    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.

  • Experimental determination of thermal contact conductance between pressure and Calandria Tubes of Indian pressurised heavy water reactors
    Nuclear Engineering and Design, 2015
    Co-Authors: A.k. Dureja, Sucharita Sinha, D. N. Pawaskar, P. Seshu, R.k. Sinha
    Abstract:

    Abstract Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a Tube type heavy water nuclear reactor, the pressure Tube is expected to sag/balloon and come in contact with outer cooler Calandria Tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two Tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure Tube and Calandria Tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  • Fitness for service assessment of coolant channels of Indian PHWRs
    Journal of Nuclear Materials, 2008
    Co-Authors: R.k. Sinha, Sucharita Sinha, K. Madhusoodanan
    Abstract:

    Abstract A typical coolant channel assembly of pressurised heavy water reactors mainly consists of pressure Tube, Calandria Tube, garter spring spacers, all made of zirconium alloys and end fittings made of SS 403. The pressure Tube is rolled at both its ends to the end fittings and is located concentrically inside the Calandria Tube with the help of garter spring spacers. Pressure Tube houses the fuel bundles, which are cooled by means of pressurised heavy water. It, thus, operates under the environment of high pressure and temperature (typically 10 MPa and 573 K), and fast neutron flux (typically 3 × 10 17  n/m 2  s, E  > 1 MeV neutrons). Under this operating environment, the material of the pressure Tube undergoes degradation over a period of time, and eventually needs to be assessed for fitness for continued operation, without jeopardising the safety of the reactor. The other components of the coolant channel assembly, which are inaccessible for any in-service inspection, are assessed for their fitness, whenever a pressure Tube is removed for either surveillance purpose or any other reasons. This paper, while describing the latest developments taking place to address the issue of fitness for service of the Zr–2.5 wt% Nb pressure Tubes, also dwells briefly upon the developments taken place, to address the issues of life management and extension of zircaloy-2 pressure Tubes in the earlier generation of Indian pressurised heavy water reactors.

Thomas W. Krause - One of the best experts on this subject based on the ideXlab platform.

  • Evaluation of Concentric Tube Model for Pressure Tube to Calandria Tube Gap Measurement
    IEEE Sensors Journal, 2019
    Co-Authors: Geoffrey Klein, Jordan E. Morelli, Mark S. Luloff, Thomas W. Krause
    Abstract:

    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.

  • Principal Components Analysis of Multifrequency Eddy Current Data Used to Measure Pressure Tube to Calandria Tube Gap
    IEEE Sensors Journal, 2016
    Co-Authors: Shaddy Shokralla, Jordan E. Morelli, Thomas W. Krause
    Abstract:

    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.

  • modelling and validation of eddy current response to changes in factors affecting pressure Tube to Calandria Tube gap measurement
    Ndt & E International, 2015
    Co-Authors: Shaddy Shokralla, Sean Sullivan, Jordan Morelli, Thomas W. Krause
    Abstract:

    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 - One of the best experts on this subject based on the ideXlab platform.

Mohammed A Quaiyum - One of the best experts on this subject based on the ideXlab platform.

  • contact conductance between cladding pressure Tube and pressure Tube Calandria Tube of advanced thermal reactor atr
    Journal of Nuclear Science and Technology, 1994
    Co-Authors: Hiroyasu Mochizuki, Mohammed A Quaiyum
    Abstract:

    In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure Tube, and the same between pressure and Calandria Tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure Tube made of Zr-2.5%Nb and Calandria Tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure Tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental re...

  • Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR)
    Journal of Nuclear Science and Technology, 1994
    Co-Authors: Hiroyasu Mochizuki, Mohammed A Quaiyum
    Abstract:

    In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure Tube, and the same between pressure and Calandria Tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure Tube made of Zr-2.5%Nb and Calandria Tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure Tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental re...

Ritu J. Singh - One of the best experts on this subject based on the ideXlab platform.

  • 3d thermo structural simulation of pressure Tube Calandria Tube behaviour under accident conditions in phwr using abaqus
    Nuclear Engineering and Design, 2018
    Co-Authors: Balbir Kumar Singh, Matthias Krause, T Nitheanandan, Ritu J. Singh, Ramesh Kumar, Avinash J. Gaikwad
    Abstract:

    Abstract In-depth understanding of the thermo-structural behaviour of structure system and components (SSCs) of a nuclear power plant is necessary for both safety margin assessment, and for devising appropriate prevention and accident mitigation strategies. In a pressurised heavy water reactor (PHWR), the fuel in the form of fuel bundles resides inside pressure Tube (PT). The pressure Tube is surrounded by a co-axial Calandria Tube (CT). The heat generated by nuclear fission in fuel bundle is carried away by heavy water coolant, which flows inside pressure Tube. The coolant is at high temperature and high pressure inside the pressure Tube. The surrounding Calandria Tubes are at low temperature and low pressure under normal operating conditions. Under certain accident scenario, the fuel bundles cooling may be lost. This results in heating up of the fuel bundles which leads to increase in temperature of the pressure Tube. The pressure Tube starts deforming as its temperature increases. The deformed pressure Tube may contact the Calandria Tube either by sagging or ballooning. This leads to increase in the Calandria Tube temperature. If the temperature on the outer surface of the Calandria Tube exceeds the critical heat flux, film boiling may occur on the surface of the Calandria Tube. If the area in dry-out is sufficiently large and the dry-out is prolonged, the pressure-Tube/Calandria-Tube combination can continue to strain radially and may challenge fuel-channel integrity. The International Collaborative Standard Problem (ICSP) on Heavy Water Reactor (HWR) Moderator Sub-cooling Requirements is organized by the IAEA to facilitate the development and validation of computer codes for the analysis of fuel channel integrity. The objective was to assess the capability of safety analysis computer codes in predicting the associated phenomena namely; radiation heat transfer from fuel to the pressure Tube (PT), from PT to Calandria Tube (CT) heat transfer, PT deformation or failure, CT to moderator heat transfer and CT deformation or failure. The current analysis is carried out as a part of international collaborative standard problem exercise. This work simulates the heat transfer phenomena along with resulting deformation of the pressure Tube and Calandria Tubes in a finite element framework. The insights obtained from the detailed modelling of the structural aspects of the phenomena can be used to update the system safety analysis codes. In this paper, a coupled heat transfer and structural analysis of pressure Tube -Calandria Tube assembly is carried out using the finite element code, ABAQUS. The heat transfer analysis implements radiation heat transfer from heater to PT and from PT to CT. Contact conductance between PT-CT is modeled based on contact pressure. Convection from outer surface of CT to water is also considered. Structural analysis included three cases elastic–creep, elasto-plastic and elasto-plastic with creep of the individual PT-CT assembly under thermal and mechanical loading. It is observed that the results matched well for the case where only elastic creep constitutive model was considered. The temperature variation of pressure Tube and Calandria Tube along with resulting deformation is obtained. The simulation is able to capture and predict the progressive contact of PT with CT and the deformation of CT along with the PT using FEM model. It was observed during the experiment and captured in the simulation that the PT-CT assembly deformation lead to removal of heat to the moderator and channel integrity was maintained for the given moderator sub cooling margin.

  • 3D-thermo-structural simulation of pressure Tube–Calandria Tube behaviour under accident conditions in PHWR using ABAQUS
    Nuclear Engineering and Design, 2018
    Co-Authors: Balbir Kumar Singh, Matthias Krause, T Nitheanandan, Ritu J. Singh, Ramesh Kumar, Avinash J. Gaikwad
    Abstract:

    In-depth understanding of the thermo-structural behaviour of structure system and components (SSCs) of a nuclear power plant is necessary for both safety margin assessment, and for devising appropriate prevention and accident mitigation strategies. In a pressurised heavy water reactor (PHWR), the fuel in the form of fuel bundles resides inside pressure Tube (PT). The pressure Tube is surrounded by a co-axial Calandria Tube (CT). The heat generated by nuclear fission in fuel bundle is carried away by heavy water coolant, which flows inside pressure Tube. The coolant is at high temperature and high pressure inside the pressure Tube. The surrounding Calandria Tubes are at low temperature and low pressure under normal operating conditions. Under certain accident scenario, the fuel bundles cooling may be lost. This results in heating up of the fuel bundles which leads to increase in temperature of the pressure Tube. The pressure Tube starts deforming as its temperature increases. The deformed pressure Tube may contact the Calandria Tube either by sagging or ballooning. This leads to increase in the Calandria Tube temperature. If the temperature on the outer surface of the Calandria Tube exceeds the critical heat flux, film boiling may occur on the surface of the Calandria Tube. If the area in dry-out is sufficiently large and the dry-out is prolonged, the pressure-Tube/Calandria-Tube combination can continue to strain radially and may challenge fuel-channel integrity. The International Collaborative Standard Problem (ICSP) on Heavy Water Reactor (HWR) Moderator Sub-cooling Requirements is organized by the IAEA to facilitate the development and validation of computer codes for the analysis of fuel channel integrity. The objective was to assess the capability of safety analysis computer codes in predicting the associated phenomena namely; radiation heat transfer from fuel to the pressure Tube (PT), from PT to Calandria Tube (CT) heat transfer, PT deformation or failure, CT to moderator heat transfer and CT deformation or failure. The current analysis is carried out as a part of international collaborative standard problem exercise. This work simulates the heat transfer phenomena along with resulting deformation of the pressure Tube and Calandria Tubes in a finite element framework. The insights obtained from the detailed modelling of the structural aspects of the phenomena can be used to update the system safety analysis codes. In this paper, a coupled heat transfer and structural analysis of pressure Tube -Calandria Tube assembly is carried out using the finite element code, ABAQUS. The heat transfer analysis implements radiation heat transfer from heater to PT and from PT to CT. Contact conductance between PT-CT is modeled based on contact pressure. Convection from outer surface of CT to water is also considered. Structural analysis included three cases elastic–creep, elasto-plastic and elasto-plastic with creep of the individual PT-CT assembly under thermal and mechanical loading. It is observed that the results matched well for the case where only elastic creep constitutive model was considered. The temperature variation of pressure Tube and Calandria Tube along with resulting deformation is obtained. The simulation is able to capture and predict the progressive contact of PT with CT and the deformation of CT along with the PT using FEM model. It was observed during the experiment and captured in the simulation that the PT-CT assembly deformation lead to removal of heat to the moderator and channel integrity was maintained for the given moderator sub cooling margin.

  • Methodology for developing channel disassembly criteria under severe accident conditions for PHWRs
    Annals of Nuclear Energy, 2011
    Co-Authors: Ritu J. Singh, K Ravi, S. K Gupta
    Abstract:

    Abstract This paper presents a methodology to develop a model for disassembly of the coolant channels in Pressurized Heavy Water Reactors under severe accident conditions. This model gives criteria to decide when under severe accident condition coolant channels will rupture due to deterioration in material properties at high temperatures and increase in load due to creep sag of channels above it and hence get disassembled. Presently available severe accident codes use simplistic and optimistic criteria based on a predefined temperature to predict failure of fuel channels and an explicit criterion for disassembly of the channel is not covered. The coolant channel disassembly model developed in this paper is based on modeling the sag and pile up of channels. A uniform temperature along the length of the channel is assumed. The disassembly of the channel is assumed when the total strain at any location exceeds the failure strain for a given temperature. A 3D failure surface which is a plot of time to failure, temperature of the Calandria Tube and load on the Calandria Tubes (on account of no of channels piled up) is developed. This failure surface can be used as an input to severe accident codes to predict the progress of the core disassembly. A set of failure surfaces is recommended to be used if metal–water reaction on the outer surface is to be accounted for loss in ductility due to metal water reaction. The temperature transient of the Calandria Tube for a severe accident obtained from system thermal hydraulic codes can be mapped onto the failure surface. The time at which the mapped transient crosses the failure surface gives the time at which the Calandria Tube is disassembled. This disassembly model is an engineered model which is much more realistic as compared to the current temperature based conservative model for predicting severe accident progression.

  • Development of core disassembly model for PHWRs under severe accidents
    2010 2nd International Conference on Reliability Safety and Hazard - Risk-Based Technologies and Physics-of-Failure Methods (ICRESH), 2010
    Co-Authors: Ritu J. Singh, K Ravi, S. K Gupta
    Abstract:

    This paper presents a model for disassembly of the Calandria Tubes in Pressurized Heavy Water Reactors (PHWR) under severe accident conditions. Such a model gives criteria to decide when under severe accident condition Calandria Tube will get disassembled. Presently available severe accident codes use simplistic criteria based on a pre defined temperature to predict disassembly of fuel channels. The coolant channel disassembly model developed in this paper is based on modeling the sag and pile up of channels. A uniform temperature distribution is assumed across the channel. The disassembly of the channel is assumed when the total strain at any location exceeds the failure strain for a given temperature. A 3D failure surface which is a plot of time to failure, temperature of the Calandria Tube and load on the Calandria Tubes (on account of no of channels piled up) is developed. This failure surface can be used as an input to severe accident codes to predict the progress of the core disassembly. The temperature transient of the Calandria Tube for a severe accident obtained from thermal hydraulic codes can be mapped onto the failure surface. The time at which the mapped transient crosses the failure surface gives the time at which the Calandria Tube is disassembled. It is also brought out that the Calandria Tube of 220 MWe Indian PHWR has enough pullout strength to avoid failure at rolled joint for the submerged channels avoiding sudden core collapse.