Anu Dutta

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.

Modeling and analysis of condensation induced water hammer

Abstract Direct contact of steam and subcooled water under certain situations may cause immense steam condensation at the two-phase interface and can lead to the generation of fast and violent pressure surges which is often termed as condensation induced water hammer (CIWH) or direct contact condensation (DCC) driven water hammer. The present work aims at the exploration of the underlying physics of the CIWH phenomenon in a horizontal two-phase flow scenario using a dedicated 1D, compressible in-house code which is formulated based on the two-fluid modeling approach (six-equation based model). The developed code is verified against the benchmark two-phase shock tube problem (Reimann problem) and it is observed that it is capable to capture the shock wave, rarefaction wave and contact discontinuity satisfactorily. A comparative assessment between present in-house code, RELAP5 and WAHA3 against the PMK-2 CIWH experimental data shows that the pressure peak amplitude predicted by our in-house code is more accurate in comparison to WAHA3 and RELAP5 simulation. In this work, emphasis is also given on the detailed investigation to study the effect of inlet water subcooling (20–80 °C), water inflow rate (corresponding  = 0.1 and 0.7) on the pressure peak amplitude (along with its occurrence time and location), phase distribution, temperature history and interfacial condensation rate during CIWH. Observation reveals that with the decrease in inlet water temperature, pressure peak magnitude increases. It is also found that the pressure peak amplitude increases with the increase in inlet water flow rate.

CFD analysis of a hydraulic valve for cavitating flow

Abstract A successful design of high pressure hydraulic valves requires a thorough analysis of both velocity and pressure fields, with the aim of improving the geometry to avoid cavitation. Cavitation behavior prediction of hydraulic valves and its associated performance drop is of high interest for the manufacturers and for the users. The paper presents a CFD analysis of the flow inside a high pressure hydraulic valve. First, the analysis was carried out without using cavitation model (single phase). It was observed that absolute pressure was going below the vapor pressure. Hence, it was required to turn on the cavitation model. This model enables formation of vapor from liquid when the pressure drops below the vaporization pressure. Since the cavitation bubble grows in a liquid at low temperature, the latent heat of evaporation can be neglected and the system can be considered isothermal. Under these conditions the pressure inside the bubble remains practically constant and the growth of the bubble radius can be approximated by the simplified Rayleigh equation. For typical poppet valve geometry, ½ of computational domain is assumed, with pressure inlet and outlet boundary conditions, and a steady flow solution is computed. Because of the highly complex geometry of the hydraulic valve, the computational domain was meshed using unstructured grids using tetrahedral cells only. The paper presents a numerical investigation of the flow inside a hydraulic valve using commercial CFD code CFD-ACE. The aim of the study is to provide a good basis for future designing of the hydraulic valve. The result indicated the cavitation zones which in turn suggest needs of modification of present geometry.

Calculation of moderator circulation in IPHWR using a porosity approach

Abstract In the present configuration of the calandria for the 700 MWe Kakrapara Nuclear Power Plant, moderator inlet diffusers are directed upwards and the outlet is from the bottom of the calandria. Moderator circulation patterns and temperature distribution needs to be predicted to ensure adequate cooling margin for all channels. This study consists of two steps: at first, an optimized calculation scheme is obtained by comparison of the predicted results with the experimental data and by evaluating the fluid flow and temperature distribution. Then, in the second step, the analysis for the real 700 MWe IPHWR moderator under normal operating conditions has been performed with the optimized scheme. The present paper describes the methodology used for predicting the circulation pattern and temperature distribution in the moderator during normal operation using CFD code CFD-ACE+. The matrix of the calandria tubes in the core region is simplified to a porous media in which the momentum resistance model is used for pressure loss. The buoyancy effects due to internal heating and jet momentum effects through inlet nozzles have been considered in the analysis. The results show that the maximum temperature observed in the calandria is within the design limits during normal operation.

Simulation model of a nuclear power plant turbine

Abstract A computer code “TURDYN” has been developed for prediction of high pressure and low pressure turbine torque under thermodynamic transient conditions. The model is based on the conservation laws of mass and energy. All the important components of turbine systems, e. g. high pressure turbine, low pressure turbine, feed heaters, reheater, moisture separator have been considered. The dynamic equations are solved simultaneously to obtain the stage pressure at various load conditions. The details of the mathematical formulation of the model and open loop responses for specific disturbances are presented.

A Computer Program for Simulating Transient Behavior in Steam Turbine Stage Pressure of AHWR

It is proposed to couple the Advanced Heavy water reactor (AHWR), which is being developed by Bhabha Atomic Research Centre, India, with a desalination plant. The objective of this coupling is to produce system make-up and domestic water. The proposed desalination plant needs about 1.9 kg/sec of steam and the minimum pressure requirement is 3 bars. The desalination plant can be fed with bled steam extracted from a suitable stage in low pressure turbine. As the turbine stage pressure changes with the load, it is essential to know the availability of bled steam at aforesaid pressure for various load condition. The objective of the present study is to identify a suitable extraction point so as to ensure availability of steam at desired condition for desalination plant, even at part load conditions. In order to fulfill the above objective a steam and feed system analysis code was developed which incorporates the mathematical formulation of different components of the steam and feed system such as, high pressure (HP) and low pressure (LP) turbines, re-heater, feed heaters etc. The dynamic equations are solved simultaneously to obtain the stage pressure at various load conditions. Based on the results obtained, the suitable extraction stage in LP turbine was selected. This enables to determine the lowest possible part load operation up to which availability of desalination plant could be ensured.Copyright