Fuel for India’s nuclear ambitions 7 April 2017

Saurav Jha gives an update on India’s plans to use its abundant thorium resources for nuclear power production.

Endowed with modest uranium reserves alongside relatively abundant thorium bearing monazite deposits, India has long pursued the ambition of setting up large scale nuclear generation capacity utilising a Th-232/U-233 cycle. The means to this end has been the three-stage nuclear programme (TNSP), which looks to sequentially build-up a sizeable fissile inventory that can support a sustainable fleet of thorium breeders. However, owing in part to its nuclear isolation during the period 1974-2008, India’s pursuit of a thorium-based programme has only yielded credible semi- industrial capability across a closed Th- 232/U-233 fuel cycle and India’s Department of Atomic Energy (DAE) currently projects large scale thorium use as being some decades away.

Nevertheless, an industrial scale demonstrator called the Advanced Heavy Water Reactor (AHWR) is projected to

be built soon and an Indian molten salt breeder reactor (IMSBR) programme has been launched to provide the mainstay

for a future thorium-based fleet. It seems climate change considerations and water desalination needs, could drive a push for earlier deployment, though this could be contingent on India’s ability to obtain fissile material internationally. DAE’s accelerator driven sub-critical system (ADSS) research and development (R&D) programme, if successful, could also provide a pathway for faster thorium-based capacity deployment.

Three stage programme

DAE’s Atomic Minerals Directorate for Exploration & Research (AMD) has to date established that India has some 12 million tonnes of monazite which contains about 1.07 million tonnes of thorium, likely the world’s greatest repository of Th-232. In contrast India has only 0.25 million tonnes of Triuranium Octoxide (U3O8) deposits. Regardless, like all others, India too had to start its nuclear programme using uranium based reactors, due to the absence of a preexisting fissile inventory. But it never gave up its strategic objective of ultimately setting up large scale thorium-based generation capacity and that is how TNSP was born.

The TNSP envisages the use of all three major fissile isotopes U-235, plutonium-239 and U-233, with only the first occurring in nature, and the other two being ‘bred’ from the fertile isotopes U-238 and Th-232, respectively. TNSP involves the setting up of both natural uranium and enriched uranium using thermal reactors in the first-stage, plutonium driven fast breeder reactors in the second stage and Th-232/U-233 cycle based ‘thermal breeders’ in the third stage. The FBRs of the second stage will be loaded with plutonium and depleted uranium from the first stage as fuel. After sufficient FBR capacity has been built up via a closed U-238/ Pu-239 cycle, Th-232 will be introduced in the blanket regions of FBRs to breed U-233. The U-233 thus bred will serve as fuel for third- stage thorium breeders.

As can be imagined, this elaborate scheme requires significant enrichment and reprocessing capability to enable a fissile build-up. Unfortunately, owing to India’s nuclear isolation post its first weapons test in 1974, the country was not able to source either equipment or fissile material that could have allowed it to meet any realistic timeline for TNSP goals. Nonetheless, India carried on thorium related research and with the end of its nuclear isolation is now looking for ways to hasten the process.

Current status

India’s thorium cycle activities though creditable, are still at a semi-industrial level. DAE has mined thousands of tonnes of monazite and extracted thorium oxide (ThO2) out of it to produce nuclear grade ThO2 powder, which in turn has been

used to fabricate fuel pins for irradiation in both power generation as well as research reactors.

Since the late 1980s, six Indian Pressurised Heavy Water Reactors (IPHWRs) have each been loaded with thirty-five 19-element ThO2 pellet bundles at different points in time.

The ThO2 bundles were irradiated up to 600 FPDs, while achieving a maximum burnup of over 13,000 Megawatt-days per ton (MWd/t). Notably, DAE says that there were no fuel failures in these experiments. The design of these ThO2 bundles was similar to the standard natural uranium (NU) bundles used in IPHWRs. These irradiation experiments were conducted not only to gain insights into reactor physics and fuel design but also for the purposes of flux flattening in the initial cores of the selected IPHWRs.

Further studies have also been conducted with fuel bundles made of both ThO2 and slightly enriched uranium (SEU) elements. DAE is currently mulling over a move to use ThO2 on a regular basis in future IPHWRs in order to build a U-233 inventory. In any case, these studies have given DAE the confidence to develop heavy-water-reactor designs that utilise large thorium loads with enriched uranium as a driver. Experiments related to Th-Pu mixed oxide (MOX) fuel for boiling water reactors have also been undertaken. Overall, DAE’s examinations seem to suggest that ThO2 based fuels exhibit better thermo- mechanical properties and slower fuel deterioration than UO2 based fuels.

Aluminium clad ‘J’ rods containing ThO2 pellets have also been irradiated in the CIRUS research reactor at the

Bhabha Atomic Research Centre (BARC), Trombay. In addition to this, Zircaloy clad test–pin assemblies made of Th-Pu MOX pellets containing 4–7% PuO2 have also been successfully irradiated in dedicated engineering loops in CIRUS to a burn-up of 18,000MWd/t without any failure. These irradiated rods have been subsequently reprocessed in DAE’s Uranium Thorium Separation Facility using the THOREX process for obtaining U-233. The U-233 obtained has been used to fabricate U-Al alloy plate fuel to drive the 30kWt KAMINI research reactor in operation at the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam since 1996. Incidentally, KAMINI is the only operating reactor in the world which uses U-233 as its primary fuel. Of late, irradiated ThO2 bundles from IPHWRs have also been reprocessed in the Power Reactor Thorium Reprocessing Facility commissioned in 2015 to obtain U-233.

In the past, ThO2 bundles have been irradiated in the blanket region of the Fast Breeder Test Reactor (FBTR) located in Kalpakkam. Moreover, design pins containing (U-233, DU, Pu) MOX meant to be used in India’s Prototype Fast Breeder Reactor (PFBR) scheduled to achieve criticality by late 2017 have also been irradiated in FBTR. Essentially, most facets of a ‘closed’ thorium based fuel cycle have been demonstrated, albeit at a semi-industrial scale.

An Industrial Scale Demonstrator

To take things to an industrial level, DAE is in the final stages of validating the design of a 300MWe (920MWt) Advanced Heavy Water Reactor (AHWR) through large scale engineering experiments in test facilities set up specifically for AHWR. Various activities related to obtaining the necessary site selection approvals and associated regulatory clearances for AHWR construction are also underway. Actual construction work on AHWR is expected to begin sometime in 2018, probably at an existing plant site such as Tarapur in India’s Maharashtra State.

In terms of design features, AHWR is a vertical pressure tube type reactor that uses boiling light water as coolant and heavy water as moderator. Its 452 coolant channels can receive both U-Th MOX and Pu-Th MOX bundles. Capable of on-power refuelling, AHWR has a design life of 100 years and uses natural circulation for core cooling and has a negative void coefficient of reactivity. AHWR’s equilibrium core will utilise Th-U-233-Pu MOX fuel in closed cycle mode. In equilibrium about 60% of the power generated by AHWR will come from U-233 bred from Th-232.

AHWR is also meant to demonstrate Indian capability in state of the art nuclear safety features, since the plan is to build thorium based reactors on the site of old semi-urban coal power plants, to eschew land and water sourcing issues. As such, the AHWR design incorporates several passive safety systems which include core cooling by natural circulation under normal operation, transients and accident scenarios. Additionally, AHWR uses passive systems for containment cooling following a loss of coolant accident. Passive safety systems have been adopted even for non-cooling applications such as a reactor trip in case of wired shut-down system failure, containment isolation and automatic depressurisation following an accident.

Interestingly, an AHWR-low enriched uranium (AHWR-LEU) variant is also being developed with a view to offering this design for export. The AHWR-LEU core will be made of Th-LEU MOX bundles and is designed to operate in once through mode. Average discharge burn-up for the AHWR-LEU is 60MWd/t as compared to 40MWd/t for the baseline AHWR. The discharged fuel from AHWR-LEU will also have much less fissile content than what is discharged by AHWR. Under 40% of the power generated by AHWR- LEU will come from U-233 bred in situ. India is offering this design as a proliferation- resistant option for countries looking to build nuclear generation capacity for the first time on their soil.

The Indian Molten Salt Breeder Reactor

Nevertheless, the staple of the third-stage of India’s TNSP is likely to be whatever is the outcome of the IMSBR programme. DAE has validated the simulation methodology it is using for various IMBSR designs by analysing and producing some of the key results of a French molten salt fast reactor concept. Early IMBSR designs rated at 850MWe of both ‘loop’ and ‘pool’ type using varied lithium- fluoride, thorium-fluoride, uranium-fluoride combinations as fuel salt and lithium-fluoride thorium-fluoride combinations as blanket salt are being explored.

DAE’s design goals will include online refuelling and reprocessing, continuous removal of gaseous fission products, high breeding ratios, low level of long-lived actinide waste generation, and large negative feedback and void coefficient of reactivity – basically the entire roster of features that are supposed to recommend MSR designs.

Incidentally, thermal neutron IMSBR designs are also being considered in addition to fast neutron designs. The IMSBR programme is expected to yield a ‘mainstay’ design for the future. The thermal breeders will have lower fissile requirements but their breeding ratios will lead to basic self- sustainment and not allow for the growth of the fleet. The fast IMSBRs on the other hand have much higher fissile requirements but could exhibit breeding ratios that may lead to a growing fleet of the type over an extended period. Obviously, for a given fissile inventory, a bigger fleet of thermal IMSBRs can be set up initially.

The critical pathway

Either way, a large U-233 inventory will have to be created, whatever the preferred IMSBR type. TSNP envisions the use of Pu-239 fuelled FBRs for this purpose wherein Th-232 will be introduced in the blanket region of FBRs to breed U-233 using neutrons released by the fast fission of Pu-239.

However, DAE intends to do this only after a large fleet of FBRs with short ‘doubling’ time has been built up. Here ‘doubling’ time refers to the time taken by a breeder reactor to generate enough fissile material that can be used to drive another such reactor. FBRs that are fuelled by plutonium and uranium in pure metallic form rather than oxide form have shorter doubling times. But metallic FBRs (MFBRs) of 1000MWe capacity are likely to be introduced only in the mid-2020s, with Th-232 blankets being put into them only in the third decade after the launch of the first such MFBR. This is why DAE does not envision significant thorium deployment before 2050.

The delayed introduction of Th-232 indicates that DAE intends to build up a large Pu-239 inventory first, as it might, since FBRs on their own are intended to give India some 200GWe of installed capacity and will be the mainstay of India’s overall reactor fleet by the mid-2030s. Nevertheless, given that India is gradually committing itself to emission targets and its peninsula requires potable water, earlier deployment of thorium is also under consideration. DAE also has thorium-based high temperature reactor (THR) designs that are suitable for water desalination purposes.

Alternatives?

TNSP’s timescales are essentially based on the use of only domestic fissile resources to build an inventory of Pu-239 and U-233 to attain a large thorium based fleet. But, if Pu- 239 were to become available from overseas, either from decommissioned weapons or by being allowed to reprocess the large stocks of accumulated spent fuel, thorium deployment could happen sooner. With India’s re-entry into global nuclear trade, there is a move underway to explore the possibility of sourcing fissile material from partner countries abroad to utilise in designs that burn plutonium while breeding U-233. Essentially India is positing thorium as a better candidate for disposing strategic fissile material compared to inert matrix carriers.

‘Imported’ Pu-239 could be ‘burned’ in emerging new designs, which though not really breeders have high ‘conversion’ ratios and will generate a lot of U-233 in lieu of the Pu-239 which is consumed. This would suit those countries that are looking to dispose of their large Pu-239 stocks due to proliferation concerns such as Japan. Of course, imported Pu-239 could also be used to set up more FBRs under ‘safeguards’, though it remains to be seen if other countries would agree to such an arrangement.

A sub (critical) route?

Simultaneously, India is also pursuing accelerator-drive system technology as an alternate pathway for thorium utilisation

as well for the transmutation of nuclear waste in a dedicated minor actinides burner reactor with ‘inherent safety against power excursions’. According to DAE, ADSS could allow faster breeding of U-233 via spallation and achieve thorium utilisation via a once through cycle.

DAE’s ADSS R&D activities are currently focused in two key areas. The first is the development of technologies that go into superconducting radio frequency (SCRF) cavity based linear accelerators (LINACs). The key technologies here include cryostats, niobium resonators, RF electronics, among others. The other area where DAE is focusing efforts is the indigenisation of previously imported equipment common to both normal and superconducting type LINACs, such as klystrons etc.

At the moment a 20MeV 30mA proton LINAC is being set up in project mode at BARC, Trombay and a high energy SCRF cavity based accelerator is set to come up in a new BARC campus in Vishakapatnam on India’s East Coast. It is difficult to project any timelines for India successfully mastering ADSS technology. However, if that were to indeed happen, India would have found for itself a potential pathway for accelerating its cherished goal of large scale thorium utilisation.