India has an ambitious nuclear power programme, but its uranium reserves are extremely limited. Total natural deposits amount to about 70 000 tons and a good proportion of this is of very poor concentration, some as low as 0.016%. By contrast, it has more than 360 000 tons of high quality thorium deposits. The benefit for India in realising this resource as part of its nuclear programme is clear.

At the inception of the nuclear power programme most members of the industry thought that to accelerate growth in the country’s installed nuclear capacity it would be essential to start building fast breeder reactors (FBRs) as quickly as possible. Uranium deposits were then estimated to be about 15 000 tons. The strategy envisioned at the time was to install as much generating capacity from pressurised heavy water reactors (PHWRs) as would be supported by the indigenous natural uranium resources and to install FBRs at a pace that could be sustained by the plutonium recovered from the irradiated natural uranium discharged from the PHWRs. Thorium was to be introduced into reactors once the installed base from the FBRs had reached a fairly large size. Eventually the large thorium deposits would sustain a U233-Th based nuclear power programme.

In the early stages, India’s nuclear community regarded scarcity of fuel as the main constraint in achieving a fast growth rate in its installed nuclear capacity. Aggressive prospecting, however, located more indigenous uranium deposits. Limited financial resources then emerged as the more stringent constraint.

The higher capital cost of FBRs made it prudent for India to spend its limited funds on installing additional PHWRs. This change in direction, with much larger natural uranium resources now available for the PHWR programme and the possibility of postponing FBR induction into the nuclear power programme, has opened the door to the use of thorium in the PHWR itself.

Self-Sustaining Equilibrium India has studied a number of fuel cycles based on thorium in a PHWR. One of the most attractive is the self-sustaining thorium cycle (SSTC). The reactor is started with natural uranium. The plutonium is recovered and recycled, along with thorium, in the same PHWR. The U-233 produced is also recycled in the same reactor. Eventually the reactor reaches the stage at which it does not require any natural uranium feed. The recovered U-233, along with thorium, is sufficient to fuel the PHWR.

Studies indicated that the total natural uranium required for putting a PHWR on the SSTC cycle is about 2.5 tons/MWe. Establishing this cycle should take about 20 years, during which period the system would gradually change from fully natural uranium in the beginning, through a thorium/plutonium stage to the final thorium-U233 stage. This cycle can also be established using thorium-U235 instead of self-generated plutonium. In this case, the natural uranium requirement is about 1.2 tons per MWe and the changeover period could be as low as six or seven years. Once it reaches the SSTC stage, the reactor (or another one in which the reactor core is identical) can be run indefinitely with thorium being the only external fuel material input. Although study results are encouraging, no steps have yet been taken to run any of the PHWRs on the SSTC cycle.

The discharge burnup of the fuel in the SSTC cycle is about 11-12 GWd/ton heavy metal. From the fuel cycle cost standpoint, this value is rather low. In order to get a higher discharge burnup, and consequently a lower fuel cycle cost, the SSTC can be topped with an external fissile material such as U-235 or plutonium. The fuel cycle will still be self-sustaining as far as U-233 is concerned. However, the higher the discharge burnup, the lower the additional energy obtained per kg of the supplementary fissile material added (see top diagram p35).

High burnup Once-Through One thorium cycle studied in detail by the West, is the once-through high burnup thorium cycle. Thorium is enriched with U-235 or plutonium to such an extent that its discharge burnup may be increased as far as the fuel can stand. Studies have been made of burnups as high as 70-90 MWd/kg.

This fuel cycle has its attractions. It has great potential for plutonium disposition, both civilian and military. Conversely, its effect is to shrink the spent fuel inventory, an advantage for countries favouring direct disposal. It is not, however, attractive from the point of view of fuel utilisation.

India has also studied various versions of this cycle. As the high cost of fuel fabrication is spread over a much larger number of kWh generated, the fuel cycle cost is low. This benefit is realised by longer residence times. But poor neutron economy results from neutron absorption in the fission products that accumulate over a long period, and also because of the higher proportion of thorium that will go to U-234 (as a result of the continued residence of protactinium 233 inside the reactor core) for an interval longer than would occur in the SSTC or other low burnup fuel cycles. The high burnup cycle does not meet India’s need for efficient fuel utilisation.

The discharge irradiation can be proportionately increased by increasing the initial plutonium content in the fuel (right). However, the residual plutonium content in the discharged fuel stays more or less constant. This cycle is thus ideally suited for plutonium disposition.

Power Flattening PHWRs achieve full power in the equilibrium core by having more highly-burnt fuel in its inner region than in the outer. This gives a certain degree of power flattening.

In the initial core less reactive fuel is loaded in the inner region.

In the first six PHWR units in India (at Rajasthan, Madras and Narora) and in all the PHWR units built in other parts of the world, the initial core contained depleted uranium bundles at selected locations in the inner region of the core. In India, apart from the availability of plenty of thorium, the possibility of obviating the need to have a campaign for the fabrication of depleted uranium fuel was an incentive to explore the possibility of using thorium fuel bundles instead of depleted uranium.

Achieving the desired degree of power flattening and ensuring the reactivity control mechanism has adequate reactivity worth throughout the period when the thorium bundles are inside the core, and in fact throughout the lifetime of the reactor, are the two objectives to be kept in mind while deciding the locations of the thorium bundles.

A dynamic programming method determined the optimal locations for the thorium bundles. The first unit of the PHWR at Kakrapar implemented the recommended thorium loading for power flattening. It was repeated in all the subsequent PHWR units that went critical. It is now the standard procedure in India to use thorium bundles for achieving power flattening in the initial PHWR core. It is worth mentioning that 35 thorium bundles can achieve the same degree of power flattening that was earlier achieved with 384 depleted uranium bundles. Some of the thorium bundles are removed during the first refuelling of the channels in which they are located, the remainder during the second refuelling operation. The last thorium bundle is removed from the reactor after about three full power years of operation.

A cross section of the core of India’s design for a 220 MWe PHWR (right) shows that it comprises 306 pressure tubes arranged along a square lattice of 22.86 cm pitch. The fuel pins and the coolant are contained within these pressure tubes. The fuel is in the form of a string of 12 bundles; each bundle is a 19-rod cluster of 49.5 cm length.

Of the 12 bundles, 10 are in the active portion of the core. The remaining two, one on each end, are outside the core. (The bundles in the active part of the core have been numbered from 1 to 10. The numbers in the 35 squares represent the location of the thorium bundle in the initial core that is optimised for power flattening without any loss in the worth of the reactivity mechanisms. For example, 1 means in that channel the thorium bundle is at one end of the channel while 10 indicates that the thorium bundle is at the other end of the channel.) Advanced Heavy Water Reactor India embarked on the design of an Advanced Heavy Water Reactor (AHWR) with two major objectives in mind: to get as much energy out of thorium as possible with the familiar heavy water reactor technology; and to eliminate the pressurised heavy water from the coolant circuit, thereby reducing the the economic cost of expensive heavy water leakage and risks associated with tritium. Safety assumed great importance in the aftermath of the Chernobyl and Three Mile Island accidents and India’s AHWR has largely met the safety goals set at this time.

The Bhabha Atomic Research Centre (BARC) has designed an AHWR with a vertical pressure tube reactor, using heavy water as moderator and boiling light water as coolant. Leaving the core, the boiling light water coolant passes up through vertical tubes that are much longer than the height of the core, before reaching the steam drum. The primary heat transport (PHT) system is designed so that the coolant flows according to the thermal siphon effect, without the help of any coolant circulating pumps. The natural convection flow is one of the characteristics of the design that makes it an inherently safe system. The BARC is demonstrating the feasibility of heat removal by natural convection using an engineering loop at the centre.

Some of the fuel assemblies in the core are thorium with self-generated U-233. The others are thorium with plutonium enrichment. The core can be optimised with different objectives – to minimise the quantity of external plutonium needed, for example, or to maximise the fraction of energy generated from thorium. Design parameters can be adjusted in order to ensure that the core has negative void and power coefficients under all conditions.

The positive void coefficient of reactivity has, of course, been the bane of many previous attempts to use boiling light water as the coolant in a heavy water reactor.

Front and Back End As early as the 1960s BARC had fabricated thorium metal rods (the so-called J rods) for irradiation in the peripheral region of the 40 MWt Cirus research reactor.

All the thorium oxide fuel bundles used in Kakrapar and subsequent PHWRs were also fabricated indigenously. The technology was developed in BARC and was transferred to the Nuclear Fuel Complex, another unit under the Indian Department of Atomic Energy. A number of thorium oxide fuel elements were supplied by India for use in the blanket of the Lotus facility in Switzerland. These fuel elements were all fabricated using powder metallurgy methods, by cold compaction and high temperature sintering.

Research and development into the fabrication of thorium fuel elements using the sol-gel method is making satisfactory progress. The technology for making sintered thorium oxide fuel and thorium metal is a mature one.

Fuel elements containing mixed oxides of thorium and uranium, or thorium and plutonium, are also being fabricated. The established powder-pellet route is being used for both (Th+U) oxide and (Th+Pu) oxide fuel. A pellet impregnation method is being tried for (Th+U) oxide fuel. A sol-gel method is being developed for thorium containing fuel. Caesium is used for simulating plutonium in the development of the sol-gel technique for (Th+Pu) mixed oxide fuel.

Before loading the thorium fuel elements in the Kakrapar core, a number of thorium fuel elements were test irradiated. Four thorium oxide fuel bundles were loaded in the Madras station in 1985. All these bundles demonstrated excellent performance characteristics. One fuel assembly containing thorium oxide and plutonium oxide was irradiated in the Cirus reactor pressurised water loop between May 1985 and January 1988. The fuel assembly was subjected to about 100 power cycles.

A facility was set up at BARC in the late sixties to reprocess the J rods discharged from Cirus in order to recover the U-233 they contained. The first lot of U-233 was separated in September 1970. A part of the U-233 was used in building the first U-233 fuelled critical facility, Purnima-2, that went critical in May 1984, using uranyl nitrate solution. A second critical facility, Purnima-3, which uses U233-Al alloy fuel was commissioned in 1990. A 30 kW reactor with U233-Al plate fuel was completed in 1996 at the Indira Gandhi Centre for Atomic Research for carrying out neutron radiography of the irradiated FBTR fuel and for activation analysis.

India’s development of the thorium cycle is now well advanced, with the nuclear community already boasting more than 30 years work. Facilities to fabricate the fuel and preprocess irradiated thorium are established and the PHWR reactor has been adapted to extract a substantial part of its power from the thorium cycle.

The groundwork for using thorium bearing fuel is underway. For India it has tremendous viability.