Condensation-Induced Water Hammer (CIWH) phenomenon involves dynamic pressure changes caused by rapid condensation of steam in subcooled water. The magnitude of pressure spike is often large and can cause damage to the pipes and components in steam-water systems. Understanding the phenomena is vital for obtaining a safe and reliable design. This paper describes experimental investigations carried out in a specially designed facility called Water and Steam Interaction Facility (WASIF-I), that offers control over water and steam flow rates, can operate in co–, cross-, or counter-current flow mode and contains multiple injection locations. The occurrence of CIWH in horizontally oriented pipes filled with saturated steam is studied for the case where subcooled water is injected in cross-flow mode. The effect of geometrical and operating variables on the initiation of CIWH as well as peak pressure are brought out in this experimental work. A correlation to predict the critical minimum feed water flowrate necessary for triggering CIWH is developed based on multiple, nonlinear regression analysis of data from around 150 experiments. A regime map is presented that can be used as a guideline to preclude occurrence of CIWH.

In Pressurized Heavy Water Reactors (PHWRs), the fuel bundles are placed in a pressure tube (PT). These bundles rest on inside of pressure tubes through the bearing pads. The pressure tubes are part of the primary pressure boundary of the reactor, and are crucial from functionality and safety point of view. Any phenomena/ factors that may cause structural degradation of PT and may affect its structural integrity, requires thorough investigation. The fretting of pressure tube at bearing pad locations is one such phenomenon, wherein the local material removal due to fretting may lead to stress concentration and associated structural degradations. Therefore, a detailed assessment of the PT fretting damage incurred during its intended life is required. Present paper covers the premise and methodology based on Archard’s wear model for evaluation of the volumetric wear rate of pressure tube. A case study has been done to evaluate the fretting wear in pressure tubes of a large Indian PHWR. 3D transient computational fluid dynamic analysis and finite element analysis with contact modelled between the fuel bundle bearing pads and pressure tube are carried out to evaluate the contact force and sliding distance time histories. Based on the analyses results, the work rate and average saturated fret depth are evaluated. The assessment revealed insignificant concern owing to the fretting of pressure tube for the considered cases.

A significant amount of hydrogen may be released inside the containment of water-cooled nuclear power reactor under postulated accident conditions. Its distribution in the multicompartment containment geometry must be known to manage and mitigate the local hydrogen concentration in combustible pockets. An experimental study to characterize the behavior of a lighter gas (helium in place of hydrogen) in a multicompartment containment studies facility (CSF) has been pursued. Helium distribution experiments have been performed in CSF by varying important accident parameters like helium release rate, injection duration, injection area, and injection direction. The experimental studies performed in CSF depict helium stratification in the upper dome region. Stratification in terms of stratification/effective stratification factor has been determined for a range of experiments. The present experimental studies are important for understanding hydrogen distribution characteristics in multicompartment containment geometry and for benchmarking computational fluid dynamics (CFD) codes. Based on these studies some important prevailing practices for recombiner placement were endorsed.

The present work reports an experimental investigation of direct contact condensation during subcooled water injection into a steam-filled horizontal pipe. The experiments have been carried out using a laboratory scale experimental facility that has been set-up in the Heat Power Laboratory of Jadavpur University. The experiments are carried out for two different initial steam pressure (2 and 3 bar) conditions and a pressure difference ranging from 1 to 3 bar between the water and steam sections. The pressure and temperature data are recorded at different salient locations within the test section. Temporal changes in temperature give an indication of water-front propagation within the test section. Observation reveals that no distinct pressure peak occurs when the pressure difference remains lowest, irrespective of the initial steam pressure within the test section. Distinct pressure peaks are, however, noticed at higher pressure differences. The amplitude of the pressure peaks is also observed to increase with the pressure difference. The maximum pressure peak amplitude, considering all the experiments, is observed to be approximately 6.08 bar at a distance of 1.51 m from the water inlet. These pressure peaks are attributed to the rapid collapse of vapour pockets when the corresponding locations become filled with subcooled water.

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 inhouse 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 degrees C), water inflow rate (corresponding Fr = 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.

Experiments were conducted using CsI aerosols in a small scale test chamber to simulate behaviour of aerosols in the containment of a nuclear reactor. The primary focus of the study was on submicron particles (14.3 nm-697.8 nm) due to their hazardous effect on human health. Different wall surfaces, viz., plexiglass, concrete and sandpaper were chosen to study the effect of surface roughness on dry deposition velocity under both quiescent and turbulent conditions. An analytical approach to calculate dry deposition velocity of submicron particles for rough surfaces has been proposed with an improvement in the existing parameterization for shift in the velocity boundary layer. The predicted deposition velocity with the improved parameterization was found to have better agreement with published measured data of Lai and Nazaroff (2005) compared to the existing parameterizations (Wood, 1981; Zhao and Wu, 2006b). There was a significant reduction in root mean square error (RMSE) between predicted, using the improved parameterization and measured deposition velocity (upto 100%) compared to earlier ones. The new analytical deposition approach was coupled with volume conserving semi-implicit coagulation model. This aerosol dynamic model was evaluated against explicit particle size distribution for the first time for rough surfaces. Normalized RMSE between simulated and measured particle size distribution varied in the range of 2%-20% at different instances. The model seems to closely predict submicron particle behaviour in indoor environment. (C) 2016 Elsevier Ltd. All rights reserved.

Direct contact condensation (DCC) is almost an inevitable phenomenon during accidental condition for all LWRs. Rapid condensation caused by the direct contact of steam and subcooled water can lead to condensation induced water hammer (CIWH). The present work explores the underlying physics of CIWH phenomenon in a horizontal pipe under different inlet conditions such as inlet water temperature, pressure difference between steam and water section, steam superheating, steam quality and duration of valve opening using RELAP5/Mod 3.4. This work emphasises on the prediction of pressure peak magnitude in conjunction with its location of occurrence under different parametric conditions. The stratified to slug flow transition is presented in terms of the 'flow regime map' which is identified as the primary cause for pressure wave generation. The strongest pressure wave amplitude due to CIWH is found to be 116.6 bar for Delta P = 10 bar. Observation reveals that peak pressure location shifts towards the subcooled water injection point for higher inlet water temperature. For the lowest inlet water temperature (T-in = 20 degrees C), the peak pressure is found at a distance of 47.5 cm away from the water inlet whereas, for the high water temperature (T-in = 120 degrees C), peak pressure is observed at 6.25 cm away from the injection point. It is also observed that the duration of valve opening significantly affects the location of peak pressure occurrence. This study also reveals that the presence of superheated or wet steam could possibly avoid the occurrence of CIWH. (C) 2016 Elsevier B.V. All rights reserved.

Containment Studies Facility (CSF) is volumetrically scaled down model of Indian Pressurized Heavy Water Reactor (IPHWR) containment for simulating LOCA/MSLB conditions which consists of concrete containment model (CM) and Primary Heat Transport Model (PHTM) vessel. Blowdown experiments at different initial vessel pressure conditions were recently conducted at CSF and the vessel and containment parameters such as pressure, temperature and level transients have been recorded during the experiments.The experimental results have been used for benchmarking of numerical procedure adopted for evaluating LOCA/MSLB conditions in nuclear containment. The numerical procedure involves simulation of blowdown phenomena using RELAP5 code for evaluating mass and energy discharge rates, which are then used for calculating containment pressure temperature transients using ASTEC and in-house CONTRAN codes. Predictions of major parameters of vessel and containment model were found to be in good agreement with that of experimental data.In containment thermal hydraulic calculations, condensation heat transfer coefficient affects the containment pressure temperature transients. Various empirical condensation models like Tagami, Uchida and Diffusion models have been incorporated in CONTRAN code and suitable condensation model has been identified for which predicted pressure values are close to the experimental one. The details of the experimental and analytical studies conducted are presented in this paper. (C) 2015 Elsevier B.V. All rights reserved.

Nuclear power plant experiences a number of transients during its operations. These transients may be due to equipment failure, malfunctioning of process support systems etc. In such a situation, the plant may result in an abnormal state which is undesired. In case of such an undesired plant condition, the operator has to carry out diagnostic and corrective actions. When an event occurs starting from the steady state operation, instruments’ readings develop a time dependent pattern and these patterns are unique with respect to the type of the particular event. Therefore, by properly selecting the plant process parameters, the transients can be distinguished. In this connection, a computer based tool known as Diagnostic and Prognostic System has been developed for identification of large pipe break scenarios in 220 MWe Pressurised Heavy Water Reactors (PHWRs) and for prediction of expected “Source Term” and consequence for a situation where Emergency Core Cooling System (ECCS) is not available or partially available. Diagnostic and Prognostic System is essentially a transient identification and expected source term forecasting system. The system is based on Artificial Neural Networks (ANNs) that continuously monitors the plant conditions and identifies a Loss Of Coolant Accident (LOCA) scenario quickly based on the reactor process parameter values. The system further identifies the availability of injection of ECCS and in case non-availability of ECCS, it can forecast expected “Source Term”. The system is a support to plant operators as well as for emergency preparedness. The ANN is trained with a process parameter database pertaining to accident conditions and tested against blind exercises. In order to see the feasibility of implementing in the plant for real-time diagnosis, this system has been set up on a high speed computing facility and has been demonstrated successfully for LOCA scenarios.

In this paper, the 2-D unsteady viscous flow around two cylinders is studied by numerical solutions of the unsteady Navier-Stokes equations with a finite element formulation. The results of a numerical investigation of the Strouhal frequencies of two identical, stationary, parallel circular cylinders arranged in staggered configurations is presented in this paper. A simple two cylinder tandem arrangement is validated for a certain range of values of spacing ratio (L/D) with few previously published results. Results of measurements of the Strouhal frequencies of circular cylinders arranged in tandem and in some selected staggered configurations are also presented. In the case of two circular cylinders, the investigation is performed at staggered angles (α) of 10°, 15°, 20° and 25° in the range of L/D =2.0 to 4.0 and Re = 100 to 15000. The findings in this study for two circular cylinders are: (i) critical spacing ratio is simulated successfully and (ii) Lift force and drag force variation for different staggered angle.

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. tinder 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, 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.

Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals. (C) 2011 Elsevier B.V. All rights reserved.

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 (LOA) 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% RIM 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 It, 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.

The objective of this study is to develop a system, which assists the operator in identifying an accident quickly using ANNs that diagnoses the accidents based on reactor process parameters, and continuously displays the status of the nuclear reactor. A large database of transient data 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—thermal-hydraulics code for the core. The present version of this system is capable of identifying large break LOCA scenarios of 220 MWe Indian PHWRs. The system has been designed to provide the necessary information to the operator to handle emergency situations when the reactor is operating. The diagnostic results obtained from ANNs study are satisfactory.

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.

Source term and their consequences have been estimated for some postulated severe accident scenarios in a 220MWe Indian Pressurised Heavy Water Reactor (IPHWR). Thermal-hydraulic analysis for the core and the containment has been studied with fission product transport within and outside the plant. The results form a data base to train a neural network based diagnostic tool useful for emergency planning and management. The paper describes in detail the assessment methodology, analyses for different events and the computed values of the doses to public.

Cross-flow around a group of cylinders is a very common phenomenon in engineering, such as flow around heat exchanger tube arrays. The cross-flow-induced vibration might cause a reduction of equipment life and might even lead to the occurrence of severe accidents. Hence, it is necessary to understand the mechanism of flowinduced vibration and the associated fluid–structure interaction in order to improve the design of such equipment. The flow of fluid behind a blunt body is difficult to compute due to the unsteady flows. The wake behind such a body consists of unordered eddies of all sizes that create large drag on the body. Over the past 30 years, a great deal of attention has been focused on research on flow around cylindrical structures, especially on flow around one or two cylinders. Nevertheless, investigations of the flow past more than two cylinders are still relatively scarce because of the numerous parameters such as geometric parameters related to cylinder arrangement, Reynolds number (Re), and boundary conditions that could affect the flow patterns. In the present paper, the flow around four cylinders in an in-line square configuration is simulated using a finite-element method based multiphysics code COMSOL. Keyword: Cross-flow, Reynolds number, Flow Induced Vibration, Strouhal number, vortex shedding