Thorium fuel cycle - potential benefits for India - IAEA publication (2005)

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India is perhaps the one of two or three nations in the world which has substantially overcome the technological and scientific challenges posed by the processes in Thorium fuel cycle. It is imperative that India builds up on this advantage to gain automatic recognition as the world's nuclear power.

Excerpts from Thorium fuel cycle - potential benefits and challenges, International Atomic Energy Agency, May, 2005
http://www-pub.iaea.org/mtcd/publications/pdf/te_1450_web.pdf

1. SUMMARY
The outlook for nuclear power around the world has generally brightened with progressive improvement in the operating performance of existing reactors, ensuring economic competitiveness of nuclear electricity in liberalized electricity markets. At the end of 2002, some 441 nuclear power plants, with total installed capacity of 358 GW(e), were in operation worldwide, generating some 16% of global electricity. In the reference scenario, the annual average rate of growth of world nuclear capacity is expected to be in the range of 0.9% up to the year 2025 by which time the total installed nuclear power would be some 438 GW(e).

Thorium is 3 to 4 times more abundant than uranium and is widely distributed in nature as an easily exploitable resource in many countries. Unlike natural uranium, which contains ~0.7% ‘fissile’ 235U isotope, natural thorium does not contain any ‘fissile’ material and is made up of the ‘fertile’ 232Th isotope only. Hence, thorium and thorium-based fuel as metal, oxide or carbide, has been utilized in combination with ‘fissile’ 235U or 239Pu in nuclear research and power reactors for conversion to ‘fissile’ 233U, thereby enlarging the ‘fissile’ material resources. During the pioneering years of nuclear energy, from the mid 1950s to mid 1970s,there was considerable interest worldwide to develop thorium fuels and fuel cycles in order to supplement uranium reserves. Thorium fuels and fuel cycles are particularly relevant to countries having large thorium deposits but very limited uranium reserves for their long term nuclear power programme. The feasibility of thorium utilization in high temperature gas cooled reactors (HTGR), light water reactors (LWR), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were demonstrated. These activities have been well documented in several extensive reviews and conference proceedings published by US Atomic Energy Commission [1], US Department of Energy [2], [3], KfA, Germany [4] and IAEA [5]. More recently, the proceedings of IAEA meetings on Thorium Fuel Utilization: Options and Trends has summarized the activities and coordinated research projects (CRP) of IAEA and the status of thorium fuel cycle option, including ADS, in Member States [9]. The initial enthusiasm on thorium fuels and fuel cycles was not sustained among the developing countries later, due to new discovery of uranium deposits and their improved availability. However, in recent times, the need for proliferation-resistance, longer fuel cycles, higher burnup, improved waste form characteristics, reduction of plutonium inventories and in situ use of bred-in fissile material has led to renewed interest in thorium-based fuels and fuel cycles in several developed countries. The two main international projects, namely Innovative Nuclear Reactors and Fuel Cycles Programme (INPRO) initiated by the IAEA and the US-led Generation IV International Forum (GIF), are also considering thorium fuels and fuel cycles. Thorium fuels and fuel cycles have the following benefits and challenges:

Benefits
• Thorium is 3 to 4 times more abundant than uranium, widely distributed in nature as an easily exploitable resource in many countries and has not been exploited commercially so far. Thorium fuels, therefore, complement uranium fuels and ensure long term sustainability of nuclear power.
• Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. In addition, the transition to thorium could be done through the incineration of weapons grade plutonium (WPu) or civilian plutonium. • The absorption cross-section for thermal neutrons of 232Th (7.4 barns) is nearly three times that of 238U (2.7 barns). Hence, a higher conversion (to 233U) is possible with 232Th than with 238U (to 239Pu). Thus, thorium is a better ‘fertile’ material than 238U in thermal reactors but thorium is inferior to depleted uranium as a ‘fertile’ material in fast reactor.
• For the ‘fissile’ 233U nuclei, the number of neutrons liberated per neutron absorbed (represented as ?) is greater than 2.0 over a wide range of thermal neutron spectrum, unlike 235U and 239Pu. Thus, contrary to 238U–239Pu cycle in which breeding can be obtained only with fast neutron spectra, the 232Th–233U fuel cycle can operate with fast, epithermal or thermal spectra.
• Thorium dioxide is chemically more stable and has higher radiation resistance than uranium dioxide. The fission product release rate for ThO2–based fuels are one order of magnitude lower than that of UO2. ThO2 has favourable thermophysical properties because of the higher thermal conductivity and lower co-efficient of thermal expansion compared to UO2. Thus, ThO2–based fuels are expected to have better in–pile performance than that of UO2 and UO2–based mixed oxide.
• ThO2 is relatively inert and does not oxidize unlike UO2, which oxidizes easily to U3O8and UO3. Hence, long term interim storage and permanent disposal in repository of spent ThO2–based fuel are simpler without the problem of oxidation.
• Th–based fuels and fuel cycles have intrinsic proliferation-resistance due to the formation of
232U via (n,2n) reactions with 232Th, 233Pa and 233U. The half-life of 232U is only 73.6 years and the daughter products have very short half-life and some like 212Bi and 208Tl emit strong gamma adiations. From the same consideration, 232U could be utilized as an attractive carrier of highly enriched uranium (HEU) and weapons grade plutonium (WPu) to avoid their proliferation for non-peaceful purpose.
• For incineration of WPu or civilian Pu in ‘once-through’ cycle, (Th, Pu)O2 fuel is more attractive, as compared to (U, Pu)O2, since plutonium is not bred in the former and the 232U formedafter the ‘once-through’ cycle in the spent fuel ensures proliferationresistance.
• In 232Th–233U fuel cycle, much lesser quantity of plutonium and long-lived Minor Actinides (MA: Np, Am and Cm) are formed as compared to the 238U–239Pu fuel cycle, thereby minimizing the radiotoxicity associated in spent fuel. However, in the back end of 232Th–233U fuel cycle, there are other radionuclides such as 231Pa, 229Th and 230U, which may have long term radiological impact.
Challenges
• The melting point of ThO2 (3 3500C) is much higher compared to that of UO2(2 8000C). Hence, a much higher sintering temperature (>2 0000C) is required to produce high density ThO2 and ThO2–based mixed oxide fuels. Admixing of ‘sintering aid’ (CaO, MgO, Nb2O5, etc) is required for achieving the desired pellet density at lower temperature.
• ThO2 and ThO2–based mixed oxide fuels are relatively inert and, unlike UO2 and (U, Pu)O2 fuels, do not dissolve easily in concentrated nitric acid. Addition of small quantities of HF in concentrated HNO3 is essential which cause corrosion of stainless steel equipment and pipings in reprocessing plants. The corrosion problem is mitigated with addition of aluminium nitrate. Boiling THOREX solution [13 M HNO3+0.05 M HF+0.1 M Al(NO3)3] at ~393 K and long dissolution period are required for ThO2–based fuels.The irradiated Th or Th–based fuels contain significant amount of 232U, which has a half-life of only 73.6 years and is associated with strong gamma emitting daughter products, 212Bi and 208Tl with very short half-life. As a result, there is significant buildup of radiation dose with storage of spent Th–based fuel or separated 233U, necessitating remote and automated reprocessing and refabrication in heavily shielded hot cells and increase in the cost of fuel cycle activities.
• In the conversion chain of 232Th to 233U, 233Pa is formed as an intermediate, which has a relatively longer half-life (~27 days) as compared to 239Np (2.35 days) in the uranium fuel cycle thereby requiring longer cooling time of at least one year for completing the decay of 233Pa to 233U. Normally, Pa is passed into the fission product waste in the THOREX process, which could have long term radiological impact. It is essential to separate Pa from the spent fuel solution prior to solvent extraction process for separation of 233U and thorium.
• The three stream process of separation of uranium, plutonium and thorium from spent (Th, Pu)O2 fuel, though viable, is yet to be developed…
Table 1 summarizes the experimental reactors and power reactors where thorium based ceramic nuclear fuels have been used in the form of ‘coated fuel particles’ (‘microspheres’) in graphite matrix in HTGRs or as Zircaloy/stainless steel clad fuel pin assemblies containing high density ‘fuel pellets’ or ‘vibratory compacted’ fuel particles or microspheres. In the past, in the two helium cooled Pebble Bed HTGRs of Germany, namely AVR 15 MW(e) and THTR 300 MW(e), ‘coated fuel particles’ of highly enriched uranium (HEU)–thorium, mixed oxide and mixed di-carbide, embedded in graphite matrix and consolidated in the form of spherical fuel elements of diameter ~60 mm were successfully utilized. Later, in the wake of international non-proliferation requirements, the HEU was replaced with low enriched uranium (LEU: <20% 235U). Coated fuel particles of mixed uranium thorium oxide and di–carbide, embedded in graphite, were also employed in the form of prismatic blocks in the helium–cooled HTGRs of USA, namely Peach Bottom (40 MW(e)) and Fort St. Vrain (330 MW(e)). The HTGR in UK, namely the Dragon reactor, has also used ‘coated fuel particles’ of mixed thorium uranium oxide and di–carbide in graphite matrix. Excerpt from Table 1. Thorium utilization in different experimental and power reactors

…In India, there has always been a strong incentive for development of thorium fuels and fuel cycles because of large thorium deposits compared to the very modest uranium reserves. Aluminium clad thorium oxide ‘pellets’ are being regularly irradiated in CIRUS and DHRUVA research reactors in BARC. Subsequently, the irradiated thoria were reprocessed by THOREX process and the recovered 233 U has been utilized in the 30kWt research reactor KAMINI in the form of Al–clad Al–20% 233U plate fuel element assemblies. Large quantities of high density sintered ThO2 pellets have been manufactured at Nuclear Fuel Complex (NFC) and are being used in: (i) fast breeder test reactor (FBTR) as stainless steel clad blanket pin assemblies and (ii) PHWRs as Zircaloy clad pin assemblies for neutron flux flattening of initial core during start–up. Several R&D activities are underway on (Th, U)O2 and (Th, Pu)O2 fuels containing <5% uranium or plutonium oxide for use in water cooled reactors and (Th, Pu)O2 containing 20–30% PuO2 and 70–80% PuO2 for use in LMFBR with large and small cores respectively. Apart from the classical ‘Powder-Pellet’ route, advanced process flowsheets, based on Sol–Gel Microsphere Pelletization (SGMP) and Impregnation techniques, amenable to automation and remotisation, have been developed for fabrication of ThO2–based mixed oxide pellets of controlled density and microstructure. Essential thermophysical properties of these fuels, including thermal conductivity, co–efficient of thermal expansion and hot hardness (in turn indentation- creep) have been evaluated. Several Zircaloy clad (Th, Pu)O2 fuel pins have been successfully irradiated to burnups in the range of 15 000–18 000 MWd/t in the pressurised water loop (PWL) of CIRUS reactor. Design and developmental activities are underway for construction of an advanced heavy water reactor of 300 MW(e) (AHWR 300) with (Th, Pu)O2 and (Th, 233U)O2 driver fuel. A case study of AHWR 300 is in progress at the IAEA, to validate the methodology finalized in Phase IA of INPRO [10].