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Avinash J. Gaikwad - 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.

  • In-vessel retention analysis for a typical PHWR
    Life Cycle Reliability and Safety Engineering, 2017
    Co-Authors: P.k. Baburajan, U. K. Paul, Avinash J. Gaikwad
    Abstract:

    A very unlikely sequence of events with multiple failures of safety systems and human actions may lead to severe accidents like simultaneous occurrence of loss of coolant accident $$\left( \text {LOCA} \right)$$ LOCA , failure of emergency core cooling system $$\left( \text {ECCS} \right)$$ ECCS and moderator cooling system. This may result in the core disassembly and debris formation in the Calandria vessel bottom. It is imperative to study the coolability/retention of the debris/corium in the Calandria vessel in the presence of the vault water as the only available heat sink. Analysis is performed with the objective to study the in-vessel retention and coolability of the corium in the Calandria vessel by external cooling through Calandria vault inventory. In-vessel retention capability with respect to the failure criterion based on failure due to reduced external cooling which is one of the six failure modes cited in the literature is investigated. Duration of the availability of the Calandria vault water for heat removal from the debris is also estimated. This study is performed using system thermal hydraulic code RELAP5 and various cases are analysed with different initial debris temperature and initial time of the transient $$\left( \text {decay power} \right)$$ decay power . Additional cases are analysed to study the effect of oxidation of $$\text {Zr}$$ Zr , porosity of the debris and the gap between the debris and Calandria vessel. It is observed that the presence of gap and porosity decreased the heat transfer and consequently increased the debris temperature. Generally, the heat flux is found to be lower than the critical heat flux values which are reported in the literature for downward facing thick vessels. The coolability of the corium is achieved in all the cases investigated as long as the external Calandria vault heat sink is available.

  • In-vessel retention analysis for a typical PHWR
    Life Cycle Reliability and Safety Engineering, 2017
    Co-Authors: P.k. Baburajan, U. K. Paul, Avinash J. Gaikwad
    Abstract:

    A very unlikely sequence of events with multiple failures of safety systems and human actions may lead to severe accidents like simultaneous occurrence of loss of coolant accident $$\left( \text {LOCA} \right)$$ , failure of emergency core cooling system $$\left( \text {ECCS} \right)$$ and moderator cooling system. This may result in the core disassembly and debris formation in the Calandria vessel bottom. It is imperative to study the coolability/retention of the debris/corium in the Calandria vessel in the presence of the vault water as the only available heat sink. Analysis is performed with the objective to study the in-vessel retention and coolability of the corium in the Calandria vessel by external cooling through Calandria vault inventory. In-vessel retention capability with respect to the failure criterion based on failure due to reduced external cooling which is one of the six failure modes cited in the literature is investigated. Duration of the availability of the Calandria vault water for heat removal from the debris is also estimated. This study is performed using system thermal hydraulic code RELAP5 and various cases are analysed with different initial debris temperature and initial time of the transient $$\left( \text {decay power} \right)$$ . Additional cases are analysed to study the effect of oxidation of $$\text {Zr}$$ , porosity of the debris and the gap between the debris and Calandria vessel. It is observed that the presence of gap and porosity decreased the heat transfer and consequently increased the debris temperature. Generally, the heat flux is found to be lower than the critical heat flux values which are reported in the literature for downward facing thick vessels. The coolability of the corium is achieved in all the cases investigated as long as the external Calandria vault heat sink is available.

  • Coupled thermo-structural analysis for in-vessel retention in PHWR using ABAQUS
    Nuclear Engineering and Design, 2017
    Co-Authors: Balbir Kumar Singh, Ritu J. Singh, Ramesh Kumar, P.k. Baburajan, Avinash J. Gaikwad
    Abstract:

    Abstract This paper presents simulations which give insights about the thermo-structural interaction of molten core debris and Calandria vessel during severe accident scenario in a pressurised heavy water reactor. The insights developed are required to devise appropriate accident mitigation strategies including in-vessel retention and source term evaluation. Internationally, the severe accident assessment research has been focussed on RPV lower head failure studies and a limited and simplified approach has been attempted for analysing Calandria vessel behaviour with core debris in pressurised heavy water reactor. This paper presents a 3D finite element based simulation of the thermo structural interaction of molten debris and Calandria vessel. During a postulated severe accident, core debris would relocate into the lower portion of Calandria vessel and may threaten the integrity of Calandria vessel. The heat transfer between the molten core debris and the vessel may cause localised overheating (or partial melting) or inelastic strain accumulation which may result in vessel failure. In this paper coupled thermo structural analysis of Calandria vessel with debris is carried out in a finite element framework to evaluate integrity of the Calandria vessel. In this analysis boundary conditions include availability and non-availability of water outside Calandria vessel. Initial conditions assume molten debris with different percentage of zirconium oxidation. Structural behaviour of Calandria vessel is simulated considering elasto-plastic material behaviour including creep deformations. A failure criterion based on inelastic strain is used in the simulation. The analysis shows that the Calandria vessel may not undergo inelastic strain failure as long as the Calandria vault water is available. The failure of Calandria vessel may occur due to the localised melting when the vault water is not available. Further the paper also presents experimental studies planned to investigate the heat transfer from Calandria with core debris to Calandria vault water. Finite element simulation is presented which is used to design the proposed direct heating arrangement to simulate the debris heat flux.

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

  • poison injection in ahwr Calandria flow pattern and mixing characteristics
    International Journal of Heat and Mass Transfer, 2017
    Co-Authors: Anuj Kumar Kansal, N K Maheshwari, Ganesh Lal Kumawat, M T Kamble, J B Joshi
    Abstract:

    Abstract In Advanced Heavy Water Reactor (AHWR), two active, independent, functionally diverse, fast acting shut down systems are used. Shut Down System-1 (SDS-1) consists of mechanical shut off rods and Shut Down System-2 (SDS-2) based on liquid poison injection into the moderator. Gadolinium nitrate solution, which acts as neutronic poison, is injected in heavy water moderator through a set of nozzles located inside Calandria vessel. This poison injection into moderator inside Calandria is expected to take place in a very short time to shut down the reactor. As SDS-2 is critical for the safety of plant, detailed knowledge of poison distribution in Calandria is required. CFD analysis using OpenFOAM has been carried out for high pressure poison injection of Shut Down System-2 of AHWR for various design options. Validation of CFD model is performed by experimental measurements. Experiments are performed for impulsively started poison jet in a 1/3.3 scaled and full-scale experimental facilities with array of tubes. In particular, the growth of poison jet is measured experimentally and CFD simulations are performed. An excellent agreement is found between CFD simulations and experimental data. CFD simulation shows that the poison covers a major portion of active core in 2 s time. In addition to this, effect of moderator flow and heat generation in moderator on poison distribution has also been studied.

  • CFD analysis of moderator flow and temperature fields inside a vertical Calandria vessel of nuclear reactor
    Nuclear Engineering and Design, 2015
    Co-Authors: Anuj Kumar Kansal, N K Maheshwari, Jyeshtharaj B. Joshi, Pallippattu Krishnan Vijayan
    Abstract:

    Abstract Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical Calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the Calandria vessel, Calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in Calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  • numerical investigation of heat transfer in the vertical annulus between pressure tube and Calandria tube of the advanced water cooled reactor
    Kerntechnik, 2010
    Co-Authors: A M Vaidya, A Borgohain, N K Maheshwari, D Govindan, P K Vijayan
    Abstract:

    Abstract In advanced water cooled reactors, an annular gap exists between pressure tube and Calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and Calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to Calandria tube. The quantification of the heat transfer between pressure tube and Calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also st...

  • Computational study of moderator flow and temperature fields in the Calandria vessel of a heavy water reactor using the PHOENICS code
    Kerntechnik, 2008
    Co-Authors: A M Vaidya, N K Maheshwari, Pallippattu Krishnan Vijayan, Dipankar Saha
    Abstract:

    Abstract Three dimensional CFD simulations of the moderator flow in the Calandria vessel of a heavy water reactor are performed using the PHOENICS CFD code. The model includes the entire Calandria vessel consisting of three shells, Calandria tubes and inlet and outlet nozzle openings. The computational model prepared in PHOENICS consists of (a) standard k-∊ turbulence model, (b) PARSOL technique for handling curved objects in cartesian grids and (c) Boussinesq formulation for handling variable density flows. PHOENICS is validated by applying it to three different flow cases. The flow pattern in the Calandria vessel under normal operating conditions is obtained through simulation. The effect of the presence of Calandria tubes and heat generation on moderator flow pattern is studied. The simulation is also performed for various heat loads and moderator mass flow rates. The maximum temperature achieved by the moderator flow under various heat loads and moderator mass flow rates is obtained.

  • Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor
    Volume 2: Fuel Cycle and High Level Waste Management; Computational Fluid Dynamics Neutronics Methods and Coupled Codes; Student Paper Competition, 2008
    Co-Authors: A M Vaidya, N K Maheshwari, Pallippattu Krishnan Vijayan, Dipankar Saha, R.k. Sinha
    Abstract:

    Computational study of the moderator flow in Calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent Calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on Calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.Copyright © 2008 by ASME

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.

  • Coupled thermo-structural analysis for in-vessel retention in PHWR using ABAQUS
    Nuclear Engineering and Design, 2017
    Co-Authors: Balbir Kumar Singh, Ritu J. Singh, Ramesh Kumar, P.k. Baburajan, Avinash J. Gaikwad
    Abstract:

    Abstract This paper presents simulations which give insights about the thermo-structural interaction of molten core debris and Calandria vessel during severe accident scenario in a pressurised heavy water reactor. The insights developed are required to devise appropriate accident mitigation strategies including in-vessel retention and source term evaluation. Internationally, the severe accident assessment research has been focussed on RPV lower head failure studies and a limited and simplified approach has been attempted for analysing Calandria vessel behaviour with core debris in pressurised heavy water reactor. This paper presents a 3D finite element based simulation of the thermo structural interaction of molten debris and Calandria vessel. During a postulated severe accident, core debris would relocate into the lower portion of Calandria vessel and may threaten the integrity of Calandria vessel. The heat transfer between the molten core debris and the vessel may cause localised overheating (or partial melting) or inelastic strain accumulation which may result in vessel failure. In this paper coupled thermo structural analysis of Calandria vessel with debris is carried out in a finite element framework to evaluate integrity of the Calandria vessel. In this analysis boundary conditions include availability and non-availability of water outside Calandria vessel. Initial conditions assume molten debris with different percentage of zirconium oxidation. Structural behaviour of Calandria vessel is simulated considering elasto-plastic material behaviour including creep deformations. A failure criterion based on inelastic strain is used in the simulation. The analysis shows that the Calandria vessel may not undergo inelastic strain failure as long as the Calandria vault water is available. The failure of Calandria vessel may occur due to the localised melting when the vault water is not available. Further the paper also presents experimental studies planned to investigate the heat transfer from Calandria with core debris to Calandria vault water. Finite element simulation is presented which is used to design the proposed direct heating arrangement to simulate the debris heat flux.

  • 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.

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

  • Retention of Molten Corium in Calandria Vessel of Phwr with Moderator Drain Pipe
    Journal of Nuclear Engineering and Radiation Science, 2020
    Co-Authors: Pradeep Pandey, Arun K. Nayak, P.p. Kulkarni, Sumit Vishnu Prasad
    Abstract:

    Abstract Retention of molten corium inside Calandria vessel is crucial for arresting accident progression in PHWRs during severe accidents. Presence of nozzles and moderator drain pipe at the bottom of Calandria vessel has not been considered in most of past studies for "In-vessel retention" (IVR). These nozzles and drain pipes used for moderator circulation can make viability of corium retention highly challenging. Once moderator has evaporated, debris reheating, compacting and melting can cause release of molten corium in moderator recirculation pipes. This can lead to relocation of corium beyond Calandria vessel. The corium might reach pump room or Calandria vault after failure of moderator cooling pipe and/or moderator pump seals. This has severe consequences on the containment integrity due to molten corium concrete interaction (MCCI). Risks posed by MCCI can be avoided if corium can be contained inside vessel even in the presence of nozzles or if at all it enters into the drain line, doesn't cause its failure. Thus it becomes crucial to evaluate the challenges faced by IVR as a severe accident management strategy due to the presence of openings in the Calandria vessel. In this paper, these issues are addressed by conducting two set of experiments; (i) with presence of debris and (ii) without debris at the bottom of CV. It was observed that debris help in arresting the molten corium front and thus prevents corium from getting into moderator cooling pipe. Without debris, molten corium was found to be relocating in the moderator drainpipe.

  • Experimental Study on Melt Coolability Capability of Calandria Vault Water During Severe Accident in Indian PHWRs for Prolonged Duration
    Journal of Nuclear Engineering and Radiation Science, 2018
    Co-Authors: Sumit Vishnu Prasad, Arun K. Nayak
    Abstract:

    The present experimental investigation in a scaled facility of an Indian pressurized heavy water reactors (PHWRs) is focused on the heat transfer behavior from the Calandria vessel (CV) to the Calandria vault during a prolonged severe accident condition in the presence of decay heat. The transient heat transfer simulates the conditions from single phase to boiling in the Calandria vault water, partial uncovery of the CV due to boil off of water in the vault, and refill of Calandria vault. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1100 °C. Decay heat in the melt pool was simulated by using high watt cartridge type heaters. The temperature distributions inside the molten pool across the CV wall thickness and vault water were measured for prolonged period which can be divided into various phases, viz., single phase natural convection heat transfer in Calandria vault, boiling heat transfer in Calandria vault, partial uncovery of CV, and refilling Calandria vault. Experimental results showed that once the crust formed, the inner vessel temperature remained very low and vessel integrity maintained. Even boiling of Calandria vault water and uncovery of CV had negligible effect on melt, CV, and vault water temperature. The heat transfer coefficients on outer vessel surface were obtained and compared with various conditions.

  • investigation of thermo mechanical behaviour in the scaled phwr stepped Calandria vessel during severe accident
    Nuclear Engineering and Design, 2017
    Co-Authors: Sumit Vishnu Prasad, A K Nayak
    Abstract:

    Abstract In case of Indian Pressurized Heavy Water Reactors (PHWRs), in-Calandria retention of corium is essential for mitigating core melt down accident. Benchmark calculations have shown that if corium breaches the Calandria vessel and enters the Calandria vault, large amount of hydrogen and other fission gases may be generated due to molten corium concrete interaction (MCCI). There is possibility of containment failure. Hence, the best option is to contain the corium inside the Calandria vessel and cool it from outside by Calandria vault water. PHWR Calandria vessel thicknesses varies from 22 mm to 32 mm and have weld joints due to the steps involved. It is true that smaller Calandria vessel thickness has lower temperature gradient and may promote higher heat transfer rate. However, there is an international concern about the healthiness of the thin vessel with stepped weld joints due to the thermal loading of high temperature corium melt which can lead to excessive thermal strain of vessel and weld joints and it may result in failure before the corium stabilization inside the vessel is achieved. In the present study, experiments were conducted in a scaled facility of an Indian PHWR to investigate the thermo-mechanical behaviour of simulated stepped Calandria vessel with weld joints. The temperature distributions and the strain contours of vessel were recorded. Numerical analysis was carried out to validate the FEM model with the test data and also analysis was extended to predict the thermo mechanical behaviour of prototypic stepped Calandria vessel.

  • In?Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues
    Journal of Nuclear Engineering and Radiation Science, 2017
    Co-Authors: Sumit Vishnu Prasad, Arun K. Nayak
    Abstract:

    After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its Calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged Calandria vessel offers inherent in-Calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the Calandria vessel by cooling it from outside using the Calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of Calandria vessel performed in scaled facilities of an Indian PHWR.

  • study on heat removal capability of Calandria vault water from molten corium in Calandria vessel during severe accident of a phwr
    Nuclear Engineering and Design, 2015
    Co-Authors: Sumit Vishnu Prasad, P K Vijayan, A K Nayak, P P Kulkarni, Keshav K Vaze
    Abstract:

    Abstract Recent Fukushima nuclear accident has triggered further awareness amongst reactor designers regarding enhancing the safety measures in a nuclear reactor. It has become important to analyse the capability of decay heat removal in a reactor to avoid radioactivity releases to the environment. Such a study has been carried out for an Indian PHWR. In a hypothetical severe core damage accident in PHWR, multiple failure of the core cooling system may lead to collapse of pressure tubes and Calandria tubes, which may ultimately relocate inside the Calandria vessel forming a debris bed. Due to decay heat generation, the debris ultimately melts down forming a molten pool inside Calandria vessel. Calandria vessel is surrounded by Calandria vault water that acts as heat sink. In order to study the extent of heat transfer from molten pool to surrounding water under severe accident condition, an experiment was carried out wherein a simulant material was poured inside a simulated Calandria vessel immersed in the simulated Calandria vault water. The amount of melt and water present in Calandria vault scaled proportionately with regard to an Indian 700 MWe PHWR. Results show that as soon as the melt was poured in the vessel, a thick crust was formed on the inner Calandria vessel, which reduced the heat transfer from the melt pool to vault water resulting high temperature gradient in melt. Even though the cylindrical vessel inner temperature was found to be very high, the water outside the vessel never boiled. When the cylindrical vessel was opened after the experiment, there was no gap observed between the vessel and crust. Numerical analysis was carried out to predict the temperature profile of the molten pool and the vessel which were in good agreement with experimental results. Results were compared for the crust growth rate and temperature profiles in the melt pool considering the decay heat and without decay heat. Results show that with no decay heat consideration, the crust thickness continuously increases with time and in case of decay heat generation, crust thickness is found to be a function of decay heat. The melt temperature is found to increase above a decay heat of 1%.

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  • cfd simulations of moderator flow inside Calandria of the passive moderator cooling system of an advanced reactor
    Nuclear Engineering and Design, 2015
    Co-Authors: Mukesh Kumar, Jyeshtharaj B. Joshi, A K Nayak, P K Vijayan
    Abstract:

    Abstract Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario ( Kumar et al., 2013 ). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria. Next, a pseudo transient simulation was carried out by taking the power and mass flow rate at different time instants to the Calandria from the 1D analysis, as an input to the 3D steady state simulations. The results indicate that temperature hotspots are located at the top center of the fuel channel matrix depending on the operating regime. Furthermore, it was observed that for a period of 7 days the temperatures are well within the safe limits.

  • study on heat removal capability of Calandria vault water from molten corium in Calandria vessel during severe accident of a phwr
    Nuclear Engineering and Design, 2015
    Co-Authors: Sumit Vishnu Prasad, P K Vijayan, A K Nayak, P P Kulkarni, Keshav K Vaze
    Abstract:

    Abstract Recent Fukushima nuclear accident has triggered further awareness amongst reactor designers regarding enhancing the safety measures in a nuclear reactor. It has become important to analyse the capability of decay heat removal in a reactor to avoid radioactivity releases to the environment. Such a study has been carried out for an Indian PHWR. In a hypothetical severe core damage accident in PHWR, multiple failure of the core cooling system may lead to collapse of pressure tubes and Calandria tubes, which may ultimately relocate inside the Calandria vessel forming a debris bed. Due to decay heat generation, the debris ultimately melts down forming a molten pool inside Calandria vessel. Calandria vessel is surrounded by Calandria vault water that acts as heat sink. In order to study the extent of heat transfer from molten pool to surrounding water under severe accident condition, an experiment was carried out wherein a simulant material was poured inside a simulated Calandria vessel immersed in the simulated Calandria vault water. The amount of melt and water present in Calandria vault scaled proportionately with regard to an Indian 700 MWe PHWR. Results show that as soon as the melt was poured in the vessel, a thick crust was formed on the inner Calandria vessel, which reduced the heat transfer from the melt pool to vault water resulting high temperature gradient in melt. Even though the cylindrical vessel inner temperature was found to be very high, the water outside the vessel never boiled. When the cylindrical vessel was opened after the experiment, there was no gap observed between the vessel and crust. Numerical analysis was carried out to predict the temperature profile of the molten pool and the vessel which were in good agreement with experimental results. Results were compared for the crust growth rate and temperature profiles in the melt pool considering the decay heat and without decay heat. Results show that with no decay heat consideration, the crust thickness continuously increases with time and in case of decay heat generation, crust thickness is found to be a function of decay heat. The melt temperature is found to increase above a decay heat of 1%.

  • thermal and structural analysis of Calandria vessel of a phwr during a severe accident
    Nuclear Engineering and Technology, 2013
    Co-Authors: P P Kulkarni, A K Nayak, Sumit Vishnu Prasad, P K Vijayan
    Abstract:

    In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and Calandria tubes, which may ultimately relocate inside the Calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the Calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the Calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the Calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the Calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the Calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

  • numerical investigation of heat transfer in the vertical annulus between pressure tube and Calandria tube of the advanced water cooled reactor
    Kerntechnik, 2010
    Co-Authors: A M Vaidya, A Borgohain, N K Maheshwari, D Govindan, P K Vijayan
    Abstract:

    Abstract In advanced water cooled reactors, an annular gap exists between pressure tube and Calandria tube. The gap is closed from top but is open from bottom. Due to differential temperature between pressure tube and Calandria tube, air flow is induced by natural convection. This leads to heat transfer from pressure tube to Calandria tube. The quantification of the heat transfer between pressure tube and Calandria tube is numerically carried out with the help of the CFD code PHOENICS. Validation of the CFD code with experimental results and some established computational work from the literature has been done in order to verify the accuracy of the code. The natural convection phenomenon in the annular gap is then simulated. The velocity and temperature fields obtained from the CFD simulation are used to compute local and average heat transfer coefficients. Heat transfer coefficients for various pressure tube temperatures are computed. The effect of water on the heat transfer in the annular gas is also st...