[en] This lecture describes the many types of phenomena encountered in the natural circulation systems of a nuclear reactor. All the components of the primary system of a nuclear reactor are described and phenomena occurring in each of them are explained. A comprehensive coverage of related thermohydraulic relationships is provided which will enable the participants to carryout the process design of a natural circulation system. (author)

[en] The main objective of this lecture is to provide deep insight into the complex natural circulation phenomena in the core of a Pressurised Heavy Water Reactor. A detailed account of natural circulation tests conducted in an Indian PHWR is given in this lecture. This will enable the participants to appreciate the importance of natural circulation in a nuclear reactor to a greater extent. (author)

[en] A theoretical model has been developed to study the local heat transfer coefficient of a condensing vapour in the presence of a noncondensable gas, where the gas/vapour mixture is flowing downward inside a vertical tube. The two-phase heat transfer is analysed using an annular flow pattern with a liquid film at the tube wall and a turbulent gas/vapour core. The gas/vapour core is modeled using the analogy between heat and mass transfer. The model incorporates Nusselt equation with McAdams modifier and Blangetti model for calculating the film heat transfer coefficient, Moody and Wallis correlations to account for film waviness effect on gas/vapour boundary layer. The suction effect due to condensation, developing flow and property variation of the gas phase is also considered. A comparative study of heat transfer coefficient and vapour mass flow rate has been made with various models to account for condensate film resistance and condensate film roughness. Results show that for very high Reynolds number, the condensation heat transfer coefficient is higher than the film heat transfer coefficient

[en] The neutrino burst from supernova 1987 detected by Mont Blanc, Kamiokande, IMB and Baksan have been studied within 0.01 ms to 1 ms by Ferraz-Mello technique to find periods from uneven spaced data. It is found that only Kamiokande and IMB neutrino data have significant periods 0.499 ms and 0.568 ms, respectively. The authors' analysis suggest that the supernova 1987A contains a submillisecond pulsar and is supported by the recently discovered submillisecond period detected from supernova 1987A optical data. This paper suggests that the pulsar in supernova 1987A is a radially pulsating neutron star

[en] For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

[en] A design optimisation of the biological shield of fast breeder test reactor was carried out using computer code HEATING. The effect of different heat sources, variation of coolant tube pitch circle radius, coolant temperature, angular pitch of coolant tubes and thermal conductivity of concrete on the temperature distribution within the shield has been studied. (author)

[en] Full text: For sustainable development of nuclear energy, a number of key issues are to be addressed. It should be economically competitive; it must address the issues related to nuclear safety, proliferation resistance, environmental impact, waste disposal and cross cutting issues like social and infra-structural aspects. To compete successfully in the long term, in the highly competitive energy market and to overcome other challenges, it is necessary to introduce innovative reactor and fuel cycle concepts. Indian Advanced Heavy Water Reactor (AHWR) is one such innovative reactor. To guide the research and development activities related to innovative concepts, user requirements are to be formulated. User requirements covering various aspects of sustainable development are being formulated at both national and international levels. One such international project involved in the formulation of user requirements is the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper deals with INPRO user requirements for safety and Indian approach to meet these requirements through AHWR

[en] The core of a boiling water reactor may see different power distributions during its operational life. How some of the typical power distributions affect some of the thermal hydraulic parameters such as pressure drop minimum critical heat flux ratio, void distribution etc. has been studied using computer code THABNA. The effect of an increase in the leakage flow has also been analysed. (author)

[en] A mathematical model has been developed to study the flow pattern transition instability which may occur in a boiling two-phase system. The model considers flow pattern transition criteria for vertical upward and horizontal flow in pipes to identify the flow pattern transition and flow pattern specific pressure drop models. It also considers the drift flux model to estimate the void fraction in the two-phase region. The model has been applied to predict the flow pattern transition instability in a natural circulation heavy water moderated boiling light water cooled reactor. It is found that the instability characteristics is similar to that of the Ledinegg-type instability. However, the number of multiple steady states for a given operating power can be much larger in the flow pattern transition instability as compared to that of the Ledinegg-type instability. Stability maps were plotted and compared for both the flow pattern transition instability and that of the Ledinegg-type instability. The influence of various geometric and operating parameters on this instability were investigated

[en] Nuclear fuel resource available in India in plentiful quantities to sustain a large power programme is thorium. At present, activities related to design and development of Advanced Heavy Water Reactor (AHWR) and Compact High Temperature (CHTR) reactor are underway for utilization of thorium. The R and D work related to these reactors is being carried out in BARC. The 300 MWe AHWR is a direct-cycle, boiling-light-water-cooled, heavy water moderated, vertical-pressure-tube-type reactor with natural circulation as mode of heat removal from core under all conditions. AHWR uses naturally available thorium as its main fuel resource, converts it into fissile ²³³U, which is burnt in-situ to generate energy. Main physics design objectives are maximization of power from Th/²³³U, negative void coefficient of reactivity, minimization of initial inventory and consumption of plutonium, self-sustaining characteristic in ²³³U and high discharge burn-up with low excess reactivity. In addition, the AHWR incorporates several passive safety features. These include core heat removal through natural circulation of the coolant in the main heat transport system, direct injection of Emergency Core Cooling System (ECCS) water in fuel, passive systems for containment cooling and isolation and availability of large inventory of water in overhead Gravity Driven Water Pool (GDWP) for continuous long-term removal of decay heat from the core. The Compact High Temperature Reactor (CHTR), a technology demonstration reactor, the design of which is underway at BARC, also uses thorium-based fuel. The coolant used in the CHTR is liquid metal lead-bismuth eutectic. This reactor also incorporates several passive systems like removal of core heat by natural circulation of liquid metal coolant in the main heat transport circuit, passive regulation and shut down systems. The reactor is also able to remove heat passively by way of conduction in the reactor block and by radiation and natural convection from the outer surface of the reactor during loss of heat sink. (author)