Vishnu Verma

Simulation of temperature distribution by finite element analysis on different components of the EXAFS beamline at INDUS-II synchrotron source

An extended X-ray absorption fine structure (EXAFS) beamline is being developed for the INDUS-II synchrotron source. Several optical and mechanical components of the beamline are exposed to high intensity synchrotron radiation while in operation. The temperature rise on different components of the beamline on exposure to the synchrotron beam has been simulated by finite element analysis. Design of the cooling mechanism for each of these components has been carried out and estimation of the temperature rise has also been done incorporating the cooling mechanism.

Thermal Analysis of Spent Fuel Transportation Cask

Spent fuel transportation casks are required to meet among others (test conditions), the regulatory thermal test conditions in order to demonstrate their ability to withstand specified accidental fire conditions during transport. This paper describes the transient thermal analysis performed with the above intention for a transportation cask. The analysis was carried out using COMSOL Multiphysics code employing the Finite Element Method (FEM). The computation covers normal transport condition, and half an hour fire test at 800°C. The objective of the analysis was to determine the maximum outer surface temperature for normal transport conditions and to assess the extent of melting of lead during fire period.

Temperature Distribution and Thermal Stress Analysis of Ball-Tank Subjected to Solar Radiation

The ball tank of the research reactor CIRUS is exposed to solar radiation. The ambient temperature undergoes seasonal and diurnal variation. The resulting thermal stress could be significant for the large structure. The temperature distribution has been obtained by the finite element method. The paper presents temperature distribution and the resulting thermal stress.

Simulated model studies for solid waste storage surveillance facility

High level radioactive waste generated from reprocessing of spent fuel from nuclear reactors are encased in canisters after vitrification. They have high heat generation rate and need interim storage under surveillance and are to be cooled continuously until major portion of the heat is dissipated. Natural circulation air cooling (using suitable stack dimensions) has been considered to cool the overpacks containing canisters. Thermal analysis has been carried out for a reduced scale model of such a facility. Theoretical and experimental results have been compared.

Dynamic Characteristics of a Hydraulic Damper

The response of a structure to earthquake or any other dynamic excitation can be brought down by using a suitable damper such as a hydraulic damper. An analytical model has been developed for hydraulic damper, which consists of a cylinder and piston arrangement with a bypass pipeline. The stiffness of the system is primarily due to the compressibility of the fluid and the damping is largely due to the pressure drop in the bypass line. The dynamic response of the hydraulic damper has been evaluated for an assumed sinusoidal motion of the piston. The paper presents detailed results of the study of the dynamic response of the damper to variations in input and system parameters. The characteristics of the damper, thus obtained, will be useful in determining the dynamic response of the whole system to which this damper will be attached.Copyright

Analysis of the CSF model for simulated loss of coolant accident conditions

Abstract An experimental facility called Containment Studies Facility (CSF) has been constructed at the Bhabha Atomic Research Centre (BARC) for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model (CM) and a Primary Heat Transport Model (PHTM) vessel. The containment model of CSF is divided into high enthalpy V1 volume (dry well) and low enthalpy V2 volume (wet well). The PHTM and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) conditions within the containment model. Experiments were carried out in CSF for 30 bar and 50 bar blowdown conditions. Pressure, temperature and other transient parameters were recorded. The experimental results were compared with outputs of modeling carried out using RELAP code and in-house containment thermal hydraulic code CONTRAN.

Preliminary evaluation of effect of Engineered Safety Features on source term for AHWR containment

Abstract The proposed Advanced Heavy Water Reactor (AHWR) employs double containment envelope along with many Engineered Safety Features (ESFs) to mitigate the consequences of Loss-of-Coolant Accidents (LOCA) with safety system failure, during which high enthalpy steam and radioactive fission products will be discharged into the containment. In such conditions, the pressurized containment will be the source of activity release to the environment by way of leakage. It is required to study the effect of ESFs on the source term from the AHWR containment. An analysis was performed to evaluate the release rate from the AHWR containment during a postulated accident with the in-house containment code CONTRAN and the aerosol behavior code NAUA5-M in a coupled way. Modules for simulating the engineered safety features were incorporated in the CONTRAN code and the aerosol transport behaviour was evaluated using NAUA5-M separately. The AHWR containment is divided into three nodal volumes interconnected by junctions. The blow down mass, energy discharge data and activity released into the containment from the reactor core, for a postulated LOCA case of 200% RIH break with failure of shutdown systems (1 & 2), are inputs to the CONTRAN code. Thermodynamic parameters like containment gas temperature, partial pressure of steam, air in the subdivided volumes along with the flow rates through junctions obtained from CONTRAN were supplied to NAUA5-M. An analysis was carried out for a number of cases, postulated based on availability/unavailability of ESFs. Pressure, temperature and activity concentration transients were evaluated, for 72 h, in the subdivided volumes along with the activity released out of the containment through leakages and stack discharges for all the cases. This paper highlights the importance of operation of ESF in reducing the activity release to the environment.

Flow accelerated corrosion study in feeder pipes

Abstract The Indian Pressurized Heavy Water Reactor (PHWR) core consists of a number of horizontal channels containing nuclear fuel bundles. Parallel coolant channels are connected to Inlet and Outlet header through feeder pipes. Coolant from Reactor Inlet Header is distributed to the coolant channels and after removing heat combines at Reactor Outlet Header. Due to space constraints the feeder pipes are joined to the channel with one or two elbows close to the end fittings of the coolant channels. The carbon steel feeder pipes carry high temperature fluid at higher velocity and are liable to undergo Flow Accelerated Corrosion (FAC). In the recent inspection it has been found that feeders having double elbow are more susceptible to FAC on the intrados of second elbow. But it was found that in some of the elbows maximum thinning due to FAC was observed on the intrados of the first elbow. Hence to resolve this, effect of first bend orientation with respect of upstream direction has been studied. Two different approaches are used for predicting the FAC rate from calculated value of wall shear stress by CFD. One method is based on evaluating of wear rate using Colburn analogy and the other using an empirical equation between wear rate and shear stress. In Colburn analogy, mass transfer coefficient is evaluated by knowing shear stress and equilibrium concentration. For a case study, wall shear stress obtained from k-∊ turbulence model was compared with k-ω SST turbulence model and no appreciable change in the wall shear stress has been found. Hence for subsequent analysis k-∊ turbulence model was chosen because large mesh size near to the surface (first layer thickness) is permitted due to higher y+ value.

Thermal analysis of VWSB-IP1 at Tarapur

Abstract High Level Liquid Radioactive Waste (HLLRW) produced during reprocessing of spent fuel from nuclear reactors is encased in the canisters after vitrification. The vitrified waste has high heat generation rate due to decay heat and needs interim storage under surveillance. The waste needs to be cooled continuously until major portion of the decay heat is dissipated. Natural circulation air cooling has been considered to cool the canisters. Canisters are placed in a storage vault and cooled by induced axial flow of air with the help of stack. The capacity of storage vault for Vitrified Waste Storage Block (VWSB) Facility proposed at Integrated Plant-1, Tarapur is designed for interim storage of waste generated of 30 yrs of IP1 plant operation. Canister and concrete temperature should be within the prescribed limit. Parametric studies have been carried out for the relevant parameters such as stack and duct dimensions, plenum height etc. Details canister temperature have been obtained using CFD code CFD-ACE+. Axial and radial temperature variation in the canisters, thimble and ventilation pipe have been evaluated in a location. Effect of natural convection (in air) within the canister and between thimble and canister is also studied. It was found that canister centerline temperature reduces by 20°C.

Neural network based diagnostic system for accident management in nuclear power plants

An artificial neural network (ANN) based diagnostic system for identification of large break loss of coolant accident (LOCA) transients in nuclear power plants (NPPs) has been developed. This is an operator support system which assists the operator in identifying a transient quickly using ANNs. A large database of transient (LOCA) analyses of reactor process parameters has been generated for reactor core, containment, environmental dispersion and radiological dose to train the ANNs. These data have been generated using various codes e.g., RELAP5, CONTRAN. The trained neural network has been integrated with the diagnostic system for future transient prediction. The main diagnostic system features several important operator support features that are useful in accident management. This paper highlights the important features of diagnostic system. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs.