Other Nuclear Facilities
The Uranium Enrichment Plant was developed in Trombay during the early 1980s while India was acquiring centrifuge technology. The facility was completed in 1985 as a pilot-scale ultracentrifuge plant. Enriched uranium from this facility was later fabricated into fuel for the CIRUS and Dhruva nuclear reactors. In 1990, a larger uranium enrichment facility, the Rare Materials Project, began operating at Mysore.
As of the early 1990s, the Uranium Enrichment Plant was operating 100 gas centrifuges. The facility reportedly has the capacity to produce two kilograms of weapons-grade uranium per year. It is not under International Atomic Energy Agency (IAEA) safeguards.
Spent Fuel Reprocessing
Operating nuclear reactors generates not just electricity but also irradiated fuel. This “spent fuel” includes uranium that has not undergone fission, plutonium, and highly radioactive materials called fission products that are produced when uranium nuclei undergo fission. BARC practices what is called reprocessing, wherein the spent fuel is dissolved in acid and various chemicals in order to separate plutonium from it. The plutonium is to be used as fuel in other reactors or to make nuclear weapons.
Reprocessing also generates multiple waste streams classified on the basis of their radioactive concentration into low, medium and high-level waste. Low-level waste has relatively low concentrations of radioactivity but comprises over 80% by volume of the wastes produced and is a major management problem. Because it is produced in such large volumes, nuclear establishments around the world routinely release low-level waste into the environment. This radioactivity makes its way into marine life. Such discharges occur routinely at Trombay.
Heavy Water production
The first heavy water plant was set up in India at Nangal in 1962. Other Heavy water plants are at Baroda, Tuticorin, Kota, Thal, Hazira Thalchar and Manuguru. The hydrogen sulphide – water process used at Kota and Manuguru plants is based on the expertise developed through indigenous R&D.
Evolution of India’s Nuclear Power Programme
|Stage 1: Pressurised Heavy Water Reactor using Natural UO2 as fuel matrix
Heavy water as moderator and coolant
Natural U isotopic composition is 0.7 % fissile U-235 and the rest is U-238. In the reactor
The first two plants were of boiling water reactors based on imported technology. Subsequent plants are of PHWR type through indigenous R&D efforts. India achieved complete self- reliance in this technology and this stage of the programme is in the industrial domain.
The future plan includes
1. Setting up of VVER type plants based on Russian Technology is under progress to augment power generation.
2. Reprocessing of spent fuel by an Open Cycle or a Closed Cycle mode.
“Open cycle” refers to disposal of the entire waste after subjecting to proper waste treatment. This Results in huge underutilization of the energy potential of Uranium (~ 2 % is exploited). “Closed cycle” refers to chemical separation of U-238 and Pu-239 and further recycled while the other radioactive fission products were separated, sorted out according to their half lives and activity and appropriately disposed off with minimum environmental disturbance.
India preferred a closed cycle mode in view of its phased expansion of nuclear power generation extending through the second and third stages.Indigenous technology for the reprocessing of the spent fuel as well as waste management programme has been developed by India through its own comprehensive R&D efforts and reprocessing plants were set up and are in operation thereby attaining self – reliance in this domain.
|STAGE 2: Fast Breeder Reactor
India’s second stage of nuclear power generation envisages the use of Pu-239 obtained from the first stage reactor operation, as the fuel core in fast breeder reactors (FBR). The main features of FBTR are
1. Pu-239 serves as the main fissile element in the FBR
2. A blanket of U-238 surrounding the fuel core will undergo nuclear transmutation to produce fresh Pu-239 as more and more Pu-239 is consumed during the operation.
3. Besides a blanket of Th-232 around the FBR core also undergoes neutron capture reactions leading to the formation of U-233. U-233 is the nuclear reactor fuel for the third stage of India’s Nuclear Power Programme.
4. It is technically feasible to produce sustained energy output of 420 GWe from FBR.
Setting up Pu-239 fuelled fast Breeder Reactor of 500 MWe power generation is in advanced stage of completion. Concurrently, it is proposed to use thorium-based fuel, along with a small feed of plutonium-based fuel in Advanced Heavy Water Reactors (AHWRs). The AHWRs are expected to shorten the period of reaching the stage of large-scale thorium utilization.
|STAGE 3: Breeder Reactor|
| The third phase of India’s Nuclear Power Generation programme is, breeder reactors using U-233 fuel. India’s vast thorium deposits permit design and operation of U-233 fuelled breeder reactors. U-233 is obtained from the nuclear transmutation of Th-232 used as a blanket in the second phase Pu-239 fuelled FBR.
Besides, U-233 fuelled breeder reactors will have a Th-232 blanket around the U-233 reactor core which will generate more U-233 as the reactor goes operational thus resulting in the production of more and more U-233 fuel from the Th-232 blanket as more of the U-233 in the fuel core is consumed helping to sustain the long term power generation fuel requirement.
These U-233/Th-232 based breeder reactors are under development and would serve as th mainstay of the final thorium utilization stage of the Indian nuclear programme. The currently known Indian thorium reserves amount to 358,000 GWe-yr of electrical energy and can easily meet the energy requirements during the next century and beyond.
Thorium is a naturally-occurring, slightly radioactive metal discovered in 1828 by the Swedish chemist Jons Jakob Berzelius, who named it after Thor, the Norse god of thunder. It is found in small amounts in most rocks and soils, where it is about three times more abundant than uranium. Soil contains an average of around 6 parts per million (ppm) of thorium. Thorium is very insoluble, which is why it is plentiful in sands but not in seawater, in contrast to uranium.
Thorium exists in nature in a single isotopic form – Th-232 – which decays very slowly (its half-life is about three times the age of the Earth). The decay chains of natural thorium and uranium give rise to minute traces of Th-228, Th-230 and Th-234, but the presence of these in mass terms is negligible. It decays eventually to lead-208.
When pure, thorium is a silvery white metal that retains its lustre for several months. However, when it is contaminated with the oxide, thorium slowly tarnishes in air, becoming grey and eventually black. When heated in air, thorium metal ignites and burns brilliantly with a white light. Thorium oxide (ThO2), also called thoria, has one of the highest melting points of all oxides (3300°C) and so it has found applications in light bulb elements, lantern mantles, arc-light lamps, welding electrodes and heat-resistant ceramics. Glass containing thorium oxide has both a high refractive index and wavelength dispersion, and is used in high quality lenses for cameras and scientific instruments.
Thorium oxide (ThO2) is relatively inert and does not oxidise further, unlike UO2. It has higher thermal conductivity and lower thermal expansion than UO2, as well as a much higher melting point. In nuclear fuel, fission gas release is much lower than in UO2.
Thorium as nuclear fuel
Thorium is a basic element of nature, like Iron and Uranium. Like Uranium, its properties allow it to be used to fuel a nuclear chain reaction that can run a power plant and make electricity (among other things). Thorium itself will not split and release energy. Rather, when it is exposed to neutrons, it will undergo a series of nuclear reactions until it eventually emerges as an isotope of uranium called U-233, which will readily split and release energy next time it absorbs a neutron. Thorium is therefore called fertile, whereas U-233 is called fissile.
Reactors that use thorium are operating on what’s called the Thorium-Uranium (Th-U) fuel cycle. The vast majority of existing or proposed nuclear reactors, however, use enriched uranium (U-235) or reprocessed plutonium (Pu-239) as fuel (in the Uranium-Plutonium cycle), and only a handful have used thorium. Current and exotic designs can theoretically accommodate thorium. The Th-U fuel cycle has some intriguing capabilities over the traditional U-Pu cycle. Of course, it has downsides as well.
Thorium cycles exclusively allow thermal breeder reactors (as opposed to fast breeders). More neutrons are released per neutron absorbed in the fuel in a traditional (thermal) type of reactor. This means that if the fuel is reprocessed, reactors could be fueled without mining any additional U-235 for reactivity boosts, which means the nuclear fuel resources on Earth can be extended by 2 orders of magnitude without some of the complications of fast reactors. Thermal breeding is perhaps best suited for Molten Salt Reactors, which are discussed on their own page as well as in summary below.
The Th-U fuel cycle does not irradiate Uranium-238 and therefore does not produce transuranic (bigger than uranium) atoms like Plutonium, Americium, Curium, etc. These transuranics are the major health concern of long-term nuclear waste. Thus, Th-U waste will be less toxic on the 10,000+ year time scale.
Kakrapar Atomic Power Station is a nuclear power station in India, which lies in the proximity of the city of Vyara in the state of Gujarat. It consists of two 220 MW pressurised water reactor with heavy water as moderator (PHWR). KAPS-1 went critical on 3 September 1992 and began commercial electricity production a few months later on 6 May 1993. KAPS-2 went critical on 8 January 1995 and began commercial production on 1 September 1995. In January 2003, CANDU Owners Group (COG) declared KAPS as the best performing pressurised heavy water reactor.