K.K. Rajan

The Alkali metal sodium in liquid state is used as the coolant in sodium cooled fast reactors because of its good nuclear properties and admirable heat transfer properties. The economic, safe and sustained operation of sodium cooled fast breeder reactors depends on reliable and dependable monitoring of process parameters such as flow, accurately in its primary, secondary and auxiliary liquid sodium circuits. Like other sensors for sodium systems the relatively good electrical conductivity of liquid sodium is successfully used for flow measurement in pipelines. Different types of electromagnetic flow meters with different configurations are designed, constructed and were successfully in use internationally. The challenges and issues faced in pipe line sodium flow measurement using electromagnetic flow meters were addressed by researchers in India and abroad. China, Russia, Japan, Korea and France have very active fast reactor programme now. SFRs are currently in operation in India, Russia and China. In India Fast Breeder Test Reactor (FBTR) is in operation and Proto Type Fast Breeder Reactor (PFBR) is being commissioned at Kalpakkam, Tamil Nadu. The size of various pipes in sodium circuits of a typical 500 MWe SFR ranges from 15 nominal bore (NB) to 800 NB. Numerical modelling and analysis of Alnico-V based flow meters, sodium calibration of the flow meters to assess the deviation from numerical prediction, stability studies on Alnico-V based flow meters, design, analysis and manufacturing of flow meter with Samarium Cobalt magnets, development of side wall type flow meters, design and development of bypass type flow meters and application of ultrasonic flow meters for sodium flow measurement are the major developments in the area of sodium flow measurement in the recent past. This paper gives a comprehensive review on the recent developments in this domain and the recommendations made for the most suited flow meter for different sizes of pipes in future SFR circuits.

Lead-Lithium (Pb-Li) eutectic alloy is considered as the coolant in Indian Lead-Lithium ceramic breeder blanket. This liquid metal alloy is circulated through the blanket modules of fusion reactors at operating temperatures around 450°C. The Lead- Lithium system which supplies the alloy to the blanket module is operating at high temperature and needs redundant level measuring sensors along with other diagnostic devices. The high operating temperature, high chemically active nature of Pb-Li alloy and its reaction with air make its instrumentation difficult over conventional instrumentation. Since Lead Lithium is a good conductor of electricity, mutual inductance type continuous level sensors is the best choice. Design and development of the level sensor has been taken up based on the experience of mutual inductance level sensor used in sodium systems. This paper provides a brief discussion about conceptual design, specifications, modeling and simulation using Finite Element Method Magnetics (FEMM), study on skin depth and temperature compensation of sensor output for the mutual induction type continuous level sensor designed for Pb-Li loop system. It also includes the experimental validation of the level probe using solid aluminum blocks in place of Lead-Lithium alloy.

Demand of Electrical Energy is increasing day by day. A steep increase in power generation capacity is needed to meet the demand. According to projections nuclear energy is having the potential to meet the demand in the long run. Twenty two nuclear power reactors are in operation now in India and have a total installed capacity of 6780 MW. Nine nuclear reactors are under construction which has a capacity of 6700 MW (including PFBR). Ten nuclear power reactors of 700 MW each are approved by Government of India recently and the site selection process is completed. Few more reactors aiming at a capacity addition of around 30000 MW are planned under international collaboration. We have limited quantity of Natural Uranium and abundant quantity of Thorium in our country. Three phase nuclear power programme of our country envisages to achieve energy security by effective use of these nuclear resources. Development and deployment of Fast Breeder Reactors with plutonium as fuel and sodium as the coolant, reprocessing of the fast reactor spent fuel to close the fuel cycle, development of advanced heavy water reactors, development of compact high temperature reactors and activities related to fusion reactor programme with international collaboration are the major advancements taking place in our country today. This paper reviews and summarizes the present status of Indian nuclear power programme and associated research and development activities progressing in our country to achieve energy security.

Electrodeionization (EDI) of cesium from cesium-sorbed ammonium molybdophosphate-polyacrylonitrile (AMP-PAN) was investigated by passing eluant through the packed bed of ion-exchange resin in an electrodialysis cell. The deionized cesium from the packed bed was recovered in catholyte by migration and in the eluant by convection. Recovery percentage of Cs by migration increased while the recovery by convection decreased with increase in current density from 20 to 40 mA/cm². Increased eluant concentration resulted in low migration percentage of cesium. Increased catholyte concentration had a negligible effect on total recovery. Apparent diffusion coefficients evaluated using the Nernst–Plank relation increased with increase in current density and catholyte concentration while a decreasing trend was observed with increase in eluant concentration.

Early detection of water or steam leaks into sodium in the steam generator units of liquid-metal-cooled fast breeder reactors is an important requirement from safety and economic considerations. Among the various methods used worldwide, one widely accepted technique is monitoring the increase in the hydrogen concentration in sodium. The increase in hydrogen concentration is due to the dissolution of the hydrogen produced during the process of sodium–water reaction. Thus any increase of H₂ from the stabilised background value indicates a leak in the steam generator. A nickel-diffuser-based instrumentation has been installed in Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, to detect the hydrogen concentration in liquid sodium. Any failure of nickel diffuser can make the whole system unavailable and hence it is required to be detected. This paper details the optimisation of design parameters for the development of a miniature mutual inductance type leak detector for the detection of sodium leak resulting from failure/rupture of nickel diffuser tubes.