# Calandria

The Experts below are selected from a list of 303 Experts worldwide ranked by ideXlab platform

### Avinash J. Gaikwad – 1st expert 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, Ritu J. Singh, Ramesh Kumar, T Nitheanandan, Matthias Krause, 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, Ritu J. Singh, Ramesh Kumar, T Nitheanandan, Matthias Krause, 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.

### N K Maheshwari – 2nd expert 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…

### Ritu J. Singh – 3rd expert 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, Ritu J. Singh, Ramesh Kumar, T Nitheanandan, Matthias Krause, 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, Ritu J. Singh, Ramesh Kumar, T Nitheanandan, Matthias Krause, 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.