India has vast reserves of Thorium, but abundance alone does not make it an energy resource. Unlocking its potential is one of the most ambitious challenges at the heart of India’s three-stage nuclear programme. Dr B S Tomar, former Director, Radiochemistry and Isotope Group at the Bhabha Atomic Research Centre, Mumbai examines the promise of Thorium, the challenges of harnessing it, and its prospects as a fuel for India’s energy future.
India follows the three-stage nuclear power programme as envisaged by Dr. Homi Jahangir Bhabha, the architect of Indian nuclear programme. The first stage is based on the pressurized heavy water reactors (PHWR) wherein natural Uranium is used as the fuel and, to sustain the neutron economy in the reactor, heavy water is used as a moderator. Heavy water is nothing but water (H2O) in which more than 99% of Hydrogen is replaced by Deuterium, the heavier isotope of Hydrogen.
Natural Uranium contains 0.72% of U-235, the fissile isotope, and 99.3% of U-238, the fertile isotope. The fissile isotope can undergo fission with thermal neutrons, while the fertile isotope cannot. It captures a thermal neutron to form the heavier isotope U-239 which undergoes beta decay to Np-239, which, in turn, undergoes beta decay to form Pu-239, the fissile isotope of Plutonium and the fuel for the second stage of the nuclear fuel cycle.
Normal water has 0.015% Deuterium. Enrichment of U-235 from 0.72% to 1-3% obviates the need for enrichment of water to heavy water (more than 99% D). The power reactors based on enriched Uranium and light water are generally Pressurised Water Reactors (PWR) and are used by several countries which have abundant sources of Uranium, and mostly use it as once through fuel.
The second stage of fuel cycle is based on the natural Uranium and Plutonium mixed fuel, in which Plutonium extracted from the spent fuel of the first stage is used. To support the second stage, India follows a closed fuel cycle wherein the spent fuel, after a cooling period of a few years, is subjected to reprocessing. The fuel reprocessing involves separation of unspent Uranium and Plutonium leaving the bulk of the fission products in the waste, which has to be managed for a long time.
The second stage of the Indian nuclear power programme received a big boost when the Prototype Fast Breeder Reactor (PFBR) achieved criticality in April 2026. The reactor uses mixed oxides of Uranium and Plutonium as fuel and liquid Sodium as coolant. There is no moderator in the reactor as it uses fast neutrons. The technology for liquid Sodium metal and mixed U-Pu fuel has been mastered using the Fast breeder test reactor (FBTR) for the last forty years. Sustaining the second stage will require an expansion of fuel reprocessing plants for separating Plutonium from the PHWR spent fuels.
The third stage is based on Thorium, which by itself is not a fissile material, but upon capture of a neutron, forms Th-233 which undergoes beta decay to Pa-233. This, in turn, undergoes beta decay to form U-233, another fissile isotope of Uranium. The fuel for the Thorium-based reactor is Thorium mixed with U-233, which is produced by irradiating Thorium in PHWRs, as is done for separating Plutonium from spent fuel.
Thorium- the fuel for sustaining energy security
India has very limited reserves of Uranium (approximately 78,000 tons of Uranium metal) and hence can provide about 420 GWe-yrs of electricity. Plutonium from PHWRs and Fast breeder reactors of second stage can provide another 54,000 GWe-yrs of electricity approximately. The vast reserves of Thorium (518,000 tons of Thorium metal) present in the beach sands of coastal states account for about one-fourth of world’s Thorium reserves, and, if harnessed successfully, can provide 358,000 GWe-yrs of electricity. That can sustain the country’s energy needs beyond the coal era. Another advantage of the Thorium-based fuels is that the spent fuel does not have long-lived minor actinides, such as, isotopes of Neptunium, Plutonium, Americium and Curium, which are present in spent fuel of Uranium-based fuels. However, as mentioned above, utilization of Thorium in India’s third stage of nuclear power programme is dependent upon surmounting the challenges, for which intense research and development (R&D) activities are going on.
Challenges in the Thorium-based nuclear fuel cycle
The U-233 separated from irradiated Thorium contains another isotope, U-232, which has daughter products emitting high energy gamma rays and hence give high radiation dose. This poses a challenge in handling such fuel pellets/rods/bundles in the conventional fuel fabrication facilities. Remote handling facilities need to be established for fabricating such a fuel, which is a daunting task. An advanced heavy water reactor (AHWR) was proposed in the early 2000s for utilizing Thorium. However, owing to the inherent challenge, the project did not fructify. The activity of these high gamma energy emitting daughter products keeps on increasing for several years and hence the option of cooling the spent fuel before reprocessing is almost ruled out. The success of the third stage of nuclear programme will, therefore, depend upon finding alternative technologies to use Thorium as a fertile material to breed U-233. Molten salt reactor (MSR) is being studied as an alternative strategy to harness the vast Thorium reserves.
Molten salt reactor (MSR) – The path forward to a Thorium-based fuel cycle
The problem of high radiation field due to U-232 in U-233 fuel can be circumvented by employing in situ breeding of U-233 from Thorium in MSRs. MSRs use a mixture of fluorides of Lithium-7 (Li-6 is a neutron poison), Beryllium and other light metals as carrier, coolant and/or moderator. The temperature attained in MSRs falls in the range of 700-800°C, which makes it a very corrosive system and hence, requires highly corrosion -resistive container material. The reactor uses low enriched U-235 to begin with and subsequently, the U-233 bred from Th-232 sustains the reactor power. The separation of fission products from the spent fuel can also be carried out in situ. The reactor has many passive safety features and hence is quite promising for utilizing Thorium in the third stage of our nuclear power programme.
Present status of MSR technology
USA: The first attempts towards developing a molten salt reactor were made in Oak Ridge National Laboratory (ORNL) of USA in the 1950s. A U-235 based molten salt reactor experiment (MSRE) was planned using Lithium-7 fluoride + Beryllium fluoride + Zirconium fluoride as the carrier salt and Uranium fluoride as fuel in 1959. The reactor power was 7.4 MWt. The reactor achieved criticality in 1965. Later in 1968, Uranium-233 fluoride was also used to demonstrate the Thorium-based molten salt breeder reactor. The experiments continued successfully till 1969 when a decision was taken to stop the experiment. However, the experiment demonstrated the feasibility of the molten salt reactor using mixed Th/U-233 as a fuel.
China: The research towards Thorium molten salt reactor (TMSR) in China began in 2011 at the Shanghai Institute of Applied Physics. During the ensuing decade, key TSMR technologies, such as molten salt production and purification were developed. The first TSMR reactor achieved criticality in October 2023. The reactor was later taken to thermal power of 2 MWt in 2024. The reactor is powered byU-235 fluoride as fuel with fluoride salts of Lithium-7 and Beryllium as carrier salts. The breeding of U-233 from Thorium was demonstrated successfully. China has a road map of demonstrating a 100MWt Thorium MSR and a vision of commercial Thorium-based MSR by 2040. TheTSMR-400 of 168 MWe is being designed as a Thorium convertor driven by low enriched Uranium.
India: Research activities towards development of molten salt technology in India were initiated at Bhabha Atomic Research Centre in the 1970s. These studies included the preparation of pure fluorides of Thorium and Lithium and solubility of Plutonium fluorides in mixture of fluorides of Lithium, Beryllium and Thorium. Some of these studies were carried out in collaboration with ORNL. However, the work was not continued. During the past decade, there has been renewed interest in the development of MSR technology. The studies include conceptual design of Molten Salt Breeder Reactor (MSBR) and R&D towards thermodynamic data on mixtures of fluorides of Th, U, Pu and fluorides of other metal ions. A Power Reactor Thoria Reprocessing Facility (PRTRF) was commissioned in 2015 at Trombay, to separate U-233 from Thoria rods irradiated in PHWRs.
Several other countries, namely, France, Germany, Netherlands, Canada, and others have also been pursuing the development of MSR technology.
The third stage of Indian nuclear power programme depends heavily on utilization of vast Thorium reserves in the country. Development of Molten salt reactor will pave the way for breeding U-233 from Thorium, which, in turn, will boost the energy security for the country far beyond the fossil fuels era. International collaboration among the different countries will accelerate the realization of Thorium-based fuel cycle in the coming decades.
