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Friday, September 27, 2013

The Fast Breeder Reactor Programme In India - Updates & Info

This presentation, below, highlights some key aspects about the Fast Breeder Reactor programme.

Slide 4 illustrates an interesting comparison, showing the additional electricity generation capability brought about by the eventual use of Thorium, in the 3rd Stage, as opposed to if India only pursues up to the second stage, or worse goes no beyond the 1st stage - realisation of the third stage, on a commercial scale, should result in a steep increase in generation potential. Critical to meeting future requirements.

Recently, the Chairman of India's Atomic Energy Commission [AEC] gave an interview where he was asked, among other things, about the present status of development of the thorium "burning" Advanced Heavy-Water Reactor [AHWR],

Well, the consultancy contract for design of conventional systems was awarded three years ago. Most of the design drawings, etc. are ready now. Our validation process is now proceeding on twin tracks. You would note that the AHWR removes core heat from the system through natural circulation (convection) of the coolant under both normal working as well as shutdown conditions, which eliminates the need for pumps driven by electrical power that we see in most other reactor designs. For this purpose, a single coolant channel of the AHWR design has been tested in BARC for a rating of up to 2.5 - 3 MWt to confirm removal of heat from the system by natural circulation.

I would point out that we are being conservative about the margins here and testing for a higher level of heat removal will be needed to exactly determine the full margin and assess if the rated power of the reactor can be accordingly raised.

To do that we are setting up a large scale AHWR thermal hydraulic test facility (ATTF), with nearly 17 MWt heating capacity for two coolant channels.

The ATTF will validate the AHWR's full potential to remove core heat through natural circulation.

The experiments will thereby also show that much greater levels of decay heat than what can possibly ensue in the event of an emergency shutdown will be removed by the AHWR through natural circulation.

In addition to this, the AHWR has many other passive safety features, including a gravity driven water pool (GDWP) containing 6600 m3 of water that can provide emergency cooling to the core. The GDWP, along with other passive safety measures means that even in a Fukushima type scenario, decay heat can be removed from the AHWR under total station blackout conditions, without availability of any external source of water or operator action for a period of 110 days at a stretch. With such features, the AHWR is considered safe enough to build a case before the regulatory authority for locating the reactor near a population centre.

The other track is, of course, proceeding at the Critical Facility located at BARC Trombay, which was commissioned in 2009 to test the reactor physics side of the design.

Now, as far as site selection for the AHWR is concerned, we haven't identified a site as of now. Given the nature of the AHWR - it is essentially a technology demonstration project- and the fact that it won't contribute a lot of power (about 300 MWe), it doesn't really make sense to have a stand-alone site for it. On the other hand, the small size of the AHWR means that it can be accommodated at an existing site, preferably close to the R&D community.


Also Read: Understanding Nuclear Energy and Technology [suggested weekend reading]