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Energy & Desalination Projects By Ratan Kumar Sinha, Bhabha Atomic Research Centre, India. 1. Is your project part of Generation IV International Forum, International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) If so, please provide details of your project’s involvement with the above organizations. India is not a member of Generation- IV International Forum. The design and development of Advanced Heavy Water Reactor (AHWR) has been carried out at Bhabha Atomic Research Centre (BARC) without any external collaboration. The IAEA’s International Project on Innovative Nuclear reactors and fuel cycles (INPRO) has stipulated a set of requirements and criteria that should be fulfilled by the innovative nuclear reactors and fuel cycles of the future. AHWR served as a case study for validating these requirements and criteria. 2. Please provide a brief description of AHWR. AHWR is a 300 MWe, vertical, pressure tube type, boiling light water cooled, and heavy water moderated reactor. The reactor incorporates a number of passive safety features and is associated with a fuel cycle having reduced environmental impact. At the same time, the reactor possesses several features, which are likely to reduce its capital and operating costs. In the Indian context, AHWR will serve as a platform for the timely development and demonstration of the reactor and fuel cycle technologies required to be in place before large scale thorium utilisation in the future. The AHWR fuel cycle has, however, enough flexibility to accommodate a large variety of fuelling options. The reactor uses thorium based oxide fuel with in-situ generated Uranium-233 and Plutonium, recovered from the spent Responses to questions by Newal Agnihotri, Editor of Nuclear Plant Journal. Ratan Kumar Sinha Mr. Ratan Kumar Sinha graduated in Mechanical Engineering in 1972 and received training in nuclear engineering, at postgraduate level, in the training school of Bhabha Atomic Research Centre (BARC) Mumbai, India. He has thirty-fi ve years of experience in the area of development of reactor engineering technologies for components and systems of pressure tube type research and power reactors. At present he is serving as Director, Reactor Design & Development Group and, Director Design, fuel of water cooled reactors, serving as fissile materials under equilibrium conditions. It addresses the requirement of sustainability of nuclear fuel resource through the use of a closed fuel cycle along with thorium. Incidentally, on account of the use of thorium based fuel, with no production of additional Plutonium and the presence of high energy gamma emitting daughter products of Uranium-232, the reactor is considered to have inherent proliferation resistant features. The production of minor actinides, in this reactor, is reduced by nearly one order of magnitude, in Manufacturing & Automation Group, BARC. His current responsibilities include directing programmes for new advanced reactors under design and development at BARC to utilise thorium. These include, the Advanced Heavy Water Reactor (AHWR), which produces most of its power from thorium, and has several innovative passive safety features. He is also responsible for the design and development of a Compact High Temperature Reactor (CHTR), which is a technology demonstrator for future Indian High Temperature Reactors intended for hydrogen generation. Mr. Sinha is a nationally and internationally recognized expert in the area of nuclear reactor technology. For the past four years he has been the Chairman of the Steering Committee of INPRO, the IAEA’s International Project on Innovative Nuclear Reactors and Fuel Cycles. Mr. Sinha has received several awards and honours. He was elected a Fellow of the Indian National Academy of Engineering in the year 1998. He has been an elected member of the Executive Committee of the Indian Nuclear Society for the last eight years. comparison with conventional reactors, thus substantially reducing the burden of managing the inventory of long-lived radioactive waste. In AHWR, light water at 259 ° C enters the core through 452 feeders, each connected to a single vertical pressure tube, in which heat is transferred from nuclear fuel leading to boiling of the coolant. The steam water mixture produced in these pressure tubes rises through tail pipes leading to four steam drums in which steam, at nominal conditions of 70 bar pressure and 270 ° C, is separated and taken to the turbine cycle. The plant is designed 22 http://subscribe.npjonline.com http://www.NPJOnline.com Nuclear Plant Journal, September-October 2008
to produce 300 MWe electricity along with 500 m 3 /day of desalinated water. The inherent and passive safety features of the reactor include negative void coefficient of reactivity, full power core heat removal using natural circulation, shut down decay heat removal backed up by natural circulation, a passive shut down device to address a postulated insider threat of disablement of the two main shut down systems, passive cooling of concrete structures surrounding the main heat transport system piping, and passive isolation of containment as well as passive cooling of containment environment following a postulated loss of coolant accident. The reactor is provided with a double containment. A 6000 m 3 capacity water tank located inside the primary containment, near its top, serves as a heat sink for a range of postulated scenarios in which the main coolant supply to the core and/or the cooling water to the condenser is not available. With the help of this heat sink and other passive features, the reactor is designed for providing a grace period of at least three days following any postulated scenario affecting the plant. Thus, even without any external source of power, coolant and operator actions, safety of the reactor is assured for practically an indefinite period. The new design features of the reactor have been validated with the help of several large experimental facilities. A large Critical Facility designed to validate the reactor physics design of AHWR has recently been commissioned at BARC. The safety related features of AHWR have been subjected to a pre-licensing design appraisal by the Indian Atomic Energy Regulatory Board. The design of the nuclear systems is nearly complete and is available for initiating the construction of the plant in the near future. 3. What has Bhabha Atomic Research Centre done in using nuclear energy in application other than power production, including District Heating, Sea Water Desalination and Transportation The Bhabha Atomic Research Centre has helped the domestic development of all required technologies, materials and hardware necessary for the Pressurised Heavy Water Reactor programme. It has also been engaged in providing the inspection and maintenance support, as needed in some critical areas for these reactors. District heating is not a major requirement in most of India with tropical climate conditions. However, the Indian programme includes 220 MWe (750 MWth) PHWRs that may be effectively deployed for a variety of applications requiring small/medium power reactors. BARC has got a very active programme in sea water desalination and its work covers a number of technologies, including membrane based technologies and evaporation based technologies for desalination and potable water production in a cost-effective as well as energy-efficient manner. The Indian Madras Atomic Power Station (MAPS), for example, is being coupled with a large desalination plant. 4. Who are Bhabha Atomic Research Centre’s partners in producing hydrogen utilizing nuclear energy Has a prototype already been tested Please include a schedule for application of hydrogen technology for transportation in India. Bhabha Atomic Research Centre has been working on the development of technologies for producing hydrogen using water splitting reactions. Its current activities in this area include conventional electrolysis, high temperature electrolysis, and chemico-thermal processes for hydrogen generation. Bhabha Atomic Research Centre is one of the several research organizations, academic institutions and industrial partners that have contributed towards preparation of a national hydrogen energy road map. 5. What is Bhabha Atomic Research Centre’s contribution to Generation IV reactors and what is the schedule for the industry to see some tangible results India is not a member of Generation- IV. However, the Advanced Heavy Water Reactor mentioned above fulfils/exceeds all the requirements stipulated by INPRO, for the next generation nuclear reactors. The design of this demonstration reactor has reached an adequately advanced level, and the construction of the reactor is planned to be initiated in the near future. 6. Who is the manufacturer of forgings for reactor pressure vessels in India The domestic Indian nuclear power programme is currently based on Pressurised Heavy Water Reactors (PHWRs) and pool type Fast Breeder Reactors (FBRs). These reactors do not require reactor pressure vessels. Major components for the Indian nuclear reactor programme have been manufactured by several industries in the governmental (public sector) as well as private sector in India. Contact: Ratan Kumar Sinha, Bhabha Atomic Research Centre, BARC, Mumbai, 400085; email: redamin@barc.gov.in. Nuclear Plant Journal, September-October 2008 http://www.NPJOnline.com http://requestinfo.npjonline.com 23
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