1st Edition
Centrifugal Pumps for Sodium Cooled Reactors
This comprehensive introduction to centrifugal pumps used in sodiumcooled fast reactors discusses the special attributes of centrifugal pumps, design features, manufacturing requirements, instrumentation, and operating experience. It covers the characteristics of mechanical pumps, used as the main coolant pumps in fast reactors.
Key Features:
- Covers description of pumps in various reactors highlighting the special features of the pumps and providing an overview of futuristic design concepts
- Discusses the aspects related to the design, manufacture, testing, instrumentation, and operating experience of centrifugal sodium pumps
- Highlights the challenges in centrifugal sodium pump testing
- Presents topics such as cavitation testing for critical applications and thermodynamic effect on pump cavitation
- Real-life case studies are included for better understanding
This book gives a detailed overview of the design, manufacture, testing, and operating experience of the main coolant pumps used in sodium-cooled nuclear reactors. It further discusses the special type of pumps used in fast reactor power plants to circulate liquid sodium through the core. The text examines the challenges in centrifugal sodium pump testing and types of test facilities around the world. Real-life examples are used to highlight important aspects. It is primarily written for senior undergraduate, graduate students, and academic researchers in the fields such as mechanical engineering, nuclear engineering, and chemical engineering.
Chapter 1
Introduction to centrifugal sodium pumps in fast reactors
Chapter 2
Description of Centrifugal Sodium Pumps
Chapter 3
Design of Centrifugal Sodium Pumps
Chapter 4
Manufacture of Centrifugal Sodium Pumps
Chapter 5
Centrifugal Sodium Pump Testing
Chapter 6
Centrifugal Sodium Pump Instrumentation
Chapter 7
Operating Experience of Reactor Centrifugal Sodium Pumps
Chapter 8
Cavitation in Centrifugal Pumps
Chapter 9
Main Coolant Pumps for Water Cooled Reactors
Chapter 10
Sodium Pumps of the Future
Appendix 1
Appendix 2
Appendix 3
Biography
Mr. R.D Kale graduated in Mechanical Engineering from the Indian Institute of Technology (IIT), Kanpur, in 1966. He completed a one-year orientation course in Nuclear Science and Engineering as part of the 10th batch of the Training school at the Bhabha Atomic Research Centre and started his career in BARC. He was deputed to CEN deCadarache, France, where he worked along with French specialists as part of the Indian design team for the design Fast Breeder Test Reactor under a collaborative project with the French atomic energy establishment from May 1969 to June 1970. He moved to the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, in 1971 and worked for a few years initially on the detailed design of some reactor components and preparation of their technical specifications. He spent the next 28 years establishing water and sodium test facilities to test fast reactor components. These included a large sodium test facility in a 43 m high building for testing full-scale reactor components in sodium and later a dedicated Steam Generator Test Facility for testing a sector prototype of the once-through sodium/water steam generator of the 500 MWe sodium cooled power plant. He spearheaded the indigenous development of sodium centrifugal pumps for the Prototype Fast Breeder Reactor (PFBR) that culminated in the manufacture of both the primary and secondary centrifugal pumps by the Indian industry. He has several publications in reputed journals, seminars, and international conferences. He has published many articles in newspapers on nuclear power and safety. He retired in the grade of Outstanding Scientist. He was the Associate Director of the Engineering Development Group and Director of the Engineering Services Group at the time of superannuation. He has also authored a book in Marathi titled "Adhunic Hindu Dharma, Kaani Kasa?" translated as "Modern Hinduism, Why and How?". In press, his recent book in Hindi is titled "Meri Yatra, 1857 kaAnkhoDekha Hal," translated into English as "My Journey, an eyewitness account of the 1857 freedom struggle". The book is a translation of the Marathi title "Maza Pravas" that traces a priest's journey from Penn Tahasil near Mumbai to Jhansi and other holy places in North India in the backdrop of tumultuous events in the first war of Indian independence in 1857.He is a member of the Institution of Engineers, India, a member of the Indian Nuclear Society, and a fellow of the prestigious Indian National Academy of Engineering (INAE).
Dr. B.K Sreedhar is in the grade of Outstanding Scientist and heads the Sodium Experiments & Hydraulics Division in the Reactor Design and Development Group at IGCAR, Kalpakkam. He graduated with a gold medal in Mechanical Engineering from the University of Calicut in 1989. He completed the year-long orientation course in Nuclear Engineering as part of the 33rd batch of the BARC Training School and joined IGCAR in 1990. He started his career in the indigenous hydraulic development of primary and secondary coolant pumps for the PFBR. He was later involved in the in-sodium testing of full-scale mechanisms/machines of PFBR. He is also engaged in development work towards realising oil-free pumps using active magnetic bearings and ferrofluid seals. He has a postgraduate degree from the Indian Institute of Technology (IIT), Madras specialising in Hydroturbomachines, and a doctorate from the Homi Bhabha National Institute (HBNI), Mumbai. He has published several papers in refereed journals and national/international seminars and conferences.He is a fellow of the Institution of Engineers, India, a member of the Indian Nuclear Society, and a fellow of the prestigious Indian National Academy of Engineering (INAE).
