India has a unique position in the world, in terms of availability of nuclear fuel resource. It has a limited resource of uranium but a large resource of thorium. The beach sands of Kerala and Orissa have rich reserves of monazite, which contains about 8 – 10% thorium.
Thorium can be used to produce nuclear energy, but not directly due to its physics characteristics. It has to be converted to 233U in a nuclear reactor, before it can be used as fuel. 233U provides better physics characteristics in comparison to the other fissile 235U and 239Pu. A Three-Stage Indian Nuclear Power Programme has been devised to utilise the available resources efficiently and in a sustainable manner.
Work on thorium has been carried out right from the inception of our nuclear programme. Studies have been carried out on all aspects of thorium fuel cycle - mining and extraction, fuel fabrication, utilisation in different reactor systems, evaluation of its various properties and irradiation behaviour, reprocessing and recycling. Some of the important milestones achieved / technological progress made in these are as follows:
The process of producing thoria from monazite is well established. IREL has produced several tonnes of nuclear grade thoria powder
The fabrication of thoria based fuel by powder-pellet method is well established. Few tonnes of thoria fuel have been fabricated at BARC and NFC for various irradiations in research and power reactors.
Studies have been carried out regarding use of thorium in different types of reactors with respect to fuel management, reactor control and fuel utilisation.
A Critical Facility has been constructed and is being used for carrying out experiments with thoria based fuels.
Post-Irradiation Examinations have been carried out on the irradiated PHWR thoria fuel bundles and (Th-Pu) MOX fuel pins.
Thermo-physical and thermodynamic properties have been evaluated for the thoria based fuels.
Thoria fuel rods irradiated in CIRUS have been reprocessed at Uranium Thorium Separation Facility (UTSF) BARC. The recovered 233U has been fabricated as fuel for KAMINI reactor.
Thoria blanket assemblies irradiated in FBTR have been reprocessed at IGCAR. The recovered 233U has been used for experimental irradiation of PFBR type fuel assembly in FBTR.
Thoria fuel bundles irradiated in PHWRs will be reprocessed in Power Reactor Thorium Reprocessing Facility (PRTRF). The recovered 233U will be used for reactor physics experiments in AHWR-Critical Facility.
Advanced reactors AHWR and AHWR300-LEU have been designed at BARC to provide impetus to the large-scale utilisation of thorium.
ADVANCE HEAVY WATER REACTOR (AHWR)
AHWR is a 300 MWe, vertical, pressure tube type, boiling light water cooled, and heavy water moderated reactor. AHWR is being set up as a technology demonstration reactor keeping in mind the long term deployment of Thorium based reactors in the third phase. It will provide a platform for demonstration of technologies required for thorium utilisation. The reactor will use (Th-Pu) MOX and (Th-233U) MOX types of fuel. The fissile 233U for this reactor will be obtained by reprocessing its spent fuel, while plutonium will be provided from reprocessing of the spent fuel of PHWRs. The adoption of closed fuel cycle in AHWR helps in generating a large fraction of energy from thorium. A co-located fuel cycle facility (FCF) is planned along with the reactor and it will have facilities for fuel fabrication, fuel reprocessing and waste management. Some of the technologically challenging issues in this are handling of the highly radioactive fresh fuel, the requirement of remote fuel fabrication and carrying three-stream aqueous reprocessing by dissolution of the stable thoria matrix.
Extensive studies on various challenges in fabrication, reprocessing and waste management of thorium fuel cycle for AHWR are being carried out at BARC.
ADVANCE HEAVY WATER REACTOR (AHWR)- LEU
AHWR-LEU is a 300 MWe, vertical, pressure tube type, boiling light water cooled, and heavy water moderated reactor. The reactor will use (Thorium-LEU) MOX as fuel with LEU (Low Enriched Uranium) having 235U enrichment of 19.75%. The reactor is being designed based on once-through fuel cycle during its life time. A provision has therefore been made for long-term storage of the spent fuel along with monitoring and retrieval. These provisions during storage will keep open the option of reprocessing the spent fuel at a later date, if required. The co-location of the fuel fabrication plant with the reactor is not essential as no recycling of the bred fissile material in the same reactor is envisaged.