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PostPosted: Jun 10, 2010 9:21 am 
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India's innovative nuclear power reactor
The Hindu, 10 June 2010
K.S.PARTHASARATHY

The reactor's development is an effort to realise futuristic objectives through innovative configuration of present-day technologies

People waiting for a nuclear renaissance expect that the new reactors on the drawing board should assure a very high level of safety and security; they must have the ability to perform with a lower level of technological infrastructure prevailing in several developing countries; they must have high fuel use efficiency and superior waste disposal options.

“The development of the Advanced Heavy Water Reactor, AHWR300-LEU, is an effort to realize these futuristic objectives through innovative configuration of present day technologies,” Anil Kakodkar and Ratan Sinha, the designers of India's innovative nuclear reactor wrote in the May 2010 issue of Nuclear Engineering International.

They called the reactor India's passive breeder.

“As a result of its fuel mix and fuel breeding properties, the 300 MWe plant requires 42 per cent less mined uranium per unit of energy produced than a modern high burn up PWR”, they added.

AHWR300-LEU with an estimated design life of 100 years is a vertical, pressure tube type, boiling light water-cooled, heavy water- moderated reactor with reduced environmental impact. It has many features which are likely to reduce both its capital and operating costs.

The designers have eliminated primary coolant pumps and drive motors and related control and power supply equipment, thereby saving the electric power to run them. This helps to reduce cost and to enhance reliability.

The use of heavy water at low pressure reduces the potential for leakages. The heat generated in the moderator will be recovered and used for heating the feed-water.

Quick replacement

The shop assembled-coolant channels have features which enable quick replacement of pressure tubes alone without affecting other components.

The design objective of the reactor is to require no exclusion zone beyond the plant boundary. The reactor will use natural circulation to remove heat from its core under operating and shut down conditions. In case the primary and the secondary shut down systems are not available due to the failure of all active systems or malicious employee action, passive injection of a “poison” — a high neutron absorbing liquid, in to the moderator will shut down the reactor.

When the reactor operates, its core will be very hot. Coolant removes the heat. If coolant is not available due to a Loss of Coolant Accident (LOCA), the emergency core cooling system (ECCS) will remove heat by passive means.

If the primary coolant tube ruptures, a large flow of water from accumulators will cool the reactor initially. Later, the core will be cooled by the injection of cold water from a 7000 cubic metre Gravity Driven Water Pool (GDWP) located at the top of the reactor building. After that, the passive containment cooling system (PCCS) provides long term containment cooling. GDWP serves as passive water sink giving a grace period of three days.

The reactor has a double containment with an elegant design which assists the formation of a passive water seal in the event of a loss of coolant accident. The seal isolates the reactor containment and the external environment, preventing the spread of radioactivity.

Fission of Uranium-233

The reactor fuel on an average contains 19.75 per cent of enriched uranium and the balance thorium oxide. A significant fraction of the reactor power, about 39 per cent, comes from the fission of Uranium-233 derived from in-situ conversion of thorium-232.
The reactor physics design has inherent safety characteristics during all conditions likely to be encountered during startup, shutdown and LOCA.

During an interview, Dr Sinha has stated that the scientists and engineers at BARC have designed a novel advanced heavy water reactor to burn thorium ( IEEE Spectrum, 2008)

“They say that because no reactor in the world today uses thorium on a large scale, they will be breaking new ground”, he added

Currently BARC has the facility for large scale validation work.

Partly as a result of this, the reactor can achieve commercial operation by 2020. Indian scientists have been exploring various fuel cycle options for improved versions of AHWR.

AHWR300-LEU has all the safety features of AHWR. It also helps in thorium utilization.

It produces much less plutonium and minor actinides compared to Pressurized Water Reactors(PWR) which is the mainstay internationally. In view of that, this reactor is more proliferation resistant.

Since minor actinides (which have relatively long half life) are less than those in PWR, it is a better choice from considerations of waste management.

AHWR300-LEU has better reactor physics characteristics.


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PostPosted: Jun 11, 2010 12:17 am 
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The fuel, according to brochure, is nearly 22% of 19.75%LEU. It also produces 39% of power from fission of Th-U233. If one is built, then it will or can be the first reactor producing a substantial part of power from thorium. KAMINI is only an experimental reactor using thorium-U233 cycle.
The author of the brochure agrees that the concept can be extended to other solid fuel reactors by just re-designing the fuel.
He has suggested a denatured fuel by mixing thorium with 19.75%LEU for the AHWR in the brochure. Proportion of LEU will be less in the PHWR and more in the LWR's. By keeping the LEU and thorium in seperate pins or bundles, U233 can be built up for thorium fueled reactors. I wish the US DOE would desist from destroying the U233 they built up some decades back.


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PostPosted: Jun 11, 2010 12:41 am 
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I have found another article on the subject reactor, but have not been able to learn whether the laser-based separation of 232U from 233U being used in this reactor.

Anyone have any details on the reprocessing used?

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PostPosted: Jun 11, 2010 1:26 am 
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GEORGELANGWORTH wrote:
I have found another article on the subject reactor, but have not been able to learn whether the laser-based separation of 232U from 233U being used in this reactor.

Anyone have any details on the reprocessing used?



http://www.ins-india.org/conf/2003/1.pdf

see – page 13 onward under

U233 cleanup

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PostPosted: Jun 11, 2010 4:22 am 
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One has'nt been built yet. The choice is still open. If the reprocessing is onsite, U233 cleaning may not be required.


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PostPosted: Jun 11, 2010 2:51 pm 
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jagdish wrote:
One has'nt been built yet. The choice is still open. If the reprocessing is onsite, U233 cleaning may not be required.



It is my current opinion that reprocessing of nuclear fuel that contains thorium into a solid fuel form absolutely requires the removal of U232 to keep nuclear workers exposed to it alive. Working with U232 requires automation. The amount of automation that is necessary to get the reprocessed fuel into the form of pellets in not economic if completely automated. It does not matter where the reprocessing is done; on-site processing notwithstanding.

With nuclear fuel in liquid form, reprocessing of nuclear fuel using automation is economically possible. This is the reason why the MSR can use any type of nuclear fuel, because purification of its fuel salt back into a liquid form is easy to do.

Simply put because of the need for automation, reprocessing of thorium based fuels into a solid form required U232 removal; reprocessing of thorium based fuels into a liquid form does not required U232 removal.

Herein underscores the requirement for the use of the Lftr in a closed thorium fuel cycle.

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PostPosted: Aug 14, 2010 10:46 am 
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http://frontierindia.net/
India to establish Global Center for Nuclear Energy Partnership, works on Thorium fuel
U233 produced in India is being used in research reactors and to prepare fuel for AHWR critical facility. Transfer of U233 with the US research labs can be best utilized in India in short term and later can be returned if required for research in the US. The research in India is well ahead but is constrained by availability of fissile feed. This problem is not there in the five recognized nuclear countries or reprocessing countries like Japan.
The fuel design of AHWR300LEU has been purposely kept as denatured for export to non-nuclear countries. If thorium and the LEU are kept in separate pins or bundles, U233 required for LFTR or any other thorium fueled reactor can be created.
Thorium reserves in India are now estimated at 850,000 tons.
Reprocessing of irradiated thorium is established in India. It is not easy for non-state actors. Even denatured Th-19.75% fuel can be reprocessed but not for weapons use and only for reactor use.


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PostPosted: Jul 15, 2011 11:41 pm 
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There is a limitation of thorium-It requires a fissile feed not available in natural thorium. There is a balancing advantage-The conversion ratio to the fissile formed in the reactor is higher at low neutron energies. Non-availability of enough fissile feed is hindering development of thorium fuel cycle in India.
The fuel design of AHWR-300LEU depends on availability of 19.75%LEU. If India goes for enrichment, as suggested in the annual report of the DAE, they should standardise it at 19.75%. It could be combined with thorium for all types of thermal reactors on the lines of this design. The returns will be:-
1. Higher electricity produced from same mined uranium.
2. Higher production of fissile isotopes including U233.
3. Higher burn up.
Russians export 20%LEU for research reactors. All research reactors running on HEU are being converted to 20%LEU. If they export it to India for fuel, a better fuel could be developed for VVER'S exported by Russia and under construction in India too.


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