D. Das

This book is introduced with a brief overview of worldwide efforts and recent outlook on the use of thoria-based fuels for power generation. It also summarizes the merits of thoria-based fuels, issues in the thorium fuel cycle, and the past to present accounts of work on the fuels. As preamble to different chapters of the book, reader’s attention is drawn toward the need of detailed information on the thermophysical, thermodynamic, and transport properties of the fuels and on the established procedures of the front and back end operations like fabrication, reprocessing, and waste management of the fuel cycle.

Over the last 30 years there has been increased interest in utilizing thorium as nuclear fuel primarily because this actinide element is three times more abundant in the earth’s crust as compared to uranium, which is widely used as nuclear fuel. With the economically extractable thorium reserve [1] in excess of 1.2 megaton its usage gives considerable savings in uranium ore and in 235U isotope enrichment units. The most abundant 232Th isotope is not fissile and its application as nuclear material rests on the fact that it absorbs slow neutron in irradiation to produce fissionable daughter nucleus 233U (232Th + n! 233Th(22m)!233

Pa(27d)!233 U(1.5x106y). With the higher neutron yield in 233U fission [2, 3] a more efficient breeding cycle compared to the cases of U or Pu can be set up. The thorium fuel cycle generates fewer long lived heavy actinides, which will be an advantage in its waste management.

For 233U breeding and for maintaining neutron economy in the core of critically run reactor, the Th-based fuels will always contain fissile components, which are usually the U or Pu isotopes in their chemically compatible forms with the host matrix. Configurationally, the sub-critical blankets mainly of thorium fuel rods surround the seed elements with highly enriched U (HEU) or Pu. The HEU/Pu seed elements and the Th blankets are spatially separated either within a given assembly, or in between assemblies of fuel rods.

D. Das (H)

Chemistry Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085, India e-mail: dasd1951@gmail. com

D. Das and S. R. Bharadwaj (eds.), Thoria-based Nuclear Fuels, Green Energy and Technology, DOI: 10.1007/978-1-4471-5589-8_1, © Springer-Verlag London 2013

The fuel bred out of thorium or a mixture of thorium and depleted uranium inherits resistance to nuclear proliferation due to the presence of the 232U isotope that decays rapidly (t1/2 = 73.6 y) to hard gamma active daughters (212Bi, 0.7-1.8 MeV, and 208Tl, 0.72-2.6 MeV) [3]. The spent fuel reprocessing or re-fabrication are not that easy as it needs elaborate remotization engineering for handling the hard gamma active fuel. This engineering blockade is helpful in manufacturing proliferation free thorium blended fuel. The Th-based fuel, (Th, Pu)O2 is thus superior for Pu-incineration as compared to (U, Pu)O2. In the absence of advanced remotization, the usual strategy of thorium-based fuel management is to go for once through cycle. Burnt fuel elements are required to be stored over decades to reduce the activity. Attaining economy in power generation using the once through strategy needs a very high discharge burnup (>60 GWd T-1). The fuel containment for achieving high burnup is an issue to be addressed for such fuel.

Many countries, particularly those rich in thorium resources, have focused attention on the research and development of the thorium-based fuels. In the past three decades the Th-based fuel cycles have been studied extensively in Germany, India, Japan, Russia, United Kingdom, and United States of America, and sig­nificant experiences have been gained on the performance of the fuel in power generation and breeding. It is generally seen that without making radical change in the configurations of the presently used power reactors like PWR, PHWR, VVER, and HTGR [3-8] or change in their operation strategies, the thorium-based fuels in oxide, alloy, or carbide forms can be used and significantly high burnup (100 GWd ton-1 or more) can be attained. Light water breeder reactor (LWBR) concept has successfully emerged [8] by using the fuel assemblies containing the seed of HEU/Pu fissile components and blankets of thorium-based material. A number of other reactor concepts have emerged with the fuel: (a) light water reactors based on the mixed oxides (Th, Pu)O2 (Pu < 5 %), (Th, U)O2 ( U/ U < 5 %), and ThO2 matrices in pellet or microsphere forms, (b) high — temperature gas-cooled reactors using SiC and pyrolytic graphite coated fuel particles of dicarbides and oxides of Th/HEU, Th/233U, and Th/Pu in pebble bed and prismatic configurations, (c) light water cooled advanced heavy water reactors (AHWR) with sub-critical core of Th/233U oxide self-sustained by a few seed regions of the conventional mixed oxide fuel (containing Pu/ U/ U < 4 %) under an overall negative void coefficient, (d) fast reactors with the mixed oxide cores, (Th, Pu/U)O2 (Pu/233U/235U * 25 %), and thoria blankets or with the alloy core, e. g., (Th? Zr? 10 % Pu), and thoria/thoria-urania blanket (e) molten salt reactors with breeding concept, and (f) accelerator driven reactor systems (ADS) employing spallation neutrons for 233U breeding in a sub-critical core of Th. Advanced CANDU reactor (ACR) is designed for operation with slightly enriched fuels (SEU) such as about 2 % enrichment for 21 GWd ton-1 burn up, or 4 % for future operation up to 45 GWd ton-1.

In USA, the investigation and utilization of thorium dioxide and thorium dioxide-uranium dioxide (thoria-urania) solid solutions as nuclear fuel materials have been successfully conducted at the Shipping port Light Water Breeder Reactor [8]. Experience with ThO2 and ThO2-UO2 fuels have been carried out at the Elk River (Minnesota) Reactor, the Indian Point (N. Y.) No. 1 Reactor, and the HTGR (High-temperature Gas-cooled Reactor) at Peach Bottom, Pennsylvania, and a commercial HTGR at Fort St. Vrain in Colorado. Recent reviews that take into consideration of pros and cons of going for thorium-based fuels in industrial scale could be seen in [3-9].

India accounting one-fourth of the world’s thorium reserve and with about six times more Th than U has aimed at the thorium utilization for large-scale energy production [9, 10]. The utilization program is being implemented through three — stage concept: 239Pu generation from uranium in pressurized heavy water reactors (PHWR), 233U breeding in 239Pu based fast breeder reactors (FBR), and 233U burning for power production. India is also developing advanced heavy water reactor (AHWR) for deriving power directly from thorium through insitu breeding of 233U. AHWR is a new concept in reactor technology. It is a vertical pressure tube type heavy water moderated reactor that has several passive safety features including the core heat removal by natural convective circulation of boiling light water. Currently, a 300 MWe reactor is being developed using the MOX fuels

ЛЛЛ ЛОА

(Th, U)O2 and (Th, 9Pu) O2 in composite cluster of 54 pins in circular array with slightly negative void coefficient of reactivity. The fissile content of the pins is kept below 4 wt%. This reactor will derive most of its power from thorium with no external input of 233U in the equilibrium cycle [10].

Many of the Th-utilization schemes for the 233U breeding and power generation involve the uses of the thoria or thoria-based fuels. Usage of the oxide matrix is primarily due to the fact that there is vast experience with oxide fuel in thermal as well as fast reactors. The performance of urania, plutonia, and their solid solutions as reactor fuel is well established. The procedures of the fuel fabrication, storage as spent fuel, reprocessing, and waste management are proven for over so many decades. The fabrication and handling of the oxides are easier than the carbide, nitride, or metallic fuels. The carbide fuel fabrication, for example, needs metic­ulous control on oxygen and moisture contents of the inert gas as carbide is highly pyrophoric and susceptible to oxidation and hydrolysis. The spent carbide fuel reprocessing is equally problematic as it is difficult to dissolve in nitric acid and the dissolution leaves behind organic complexes. The uses of carbide, nitride, or metallic alloy fuels are generally considered as advanced concepts to cater the strategic need of compact reactor core to achieve high breeding gain, and dispo­sition of weapon grade Pu.

Like the case of the conventionally used oxide fuels, the thoria-based fuels do not pose any difficulty in handling and fabrication in the virgin state. Thoria does not get oxidized or easily hydrolyzed, and as compared to urania it has better chemical stability and desirable thermophysical and radiation resistance properties which ensures better in-pile performance and a more stable waste form. However, as mentioned, the thoria-based spent fuel handling is exceptionally difficult owing to the presence of hard gamma emitting nuclei. Under such situation one adopts extensive burning inside reactor so that the overall economy in the Th-utilization program is met in the once through cycle. Based on the success of attaining high burnup (50-100 GWd ton-1) in the exploratory runs in experimental reactors, the present target is to achieve the same on commercial basis in PWR/PHWR and FBR configurations.

For designing nuclear reactors based on thoria-based fuels it is necessary to have a thorough analysis of the fuel performance using proven simulation code and reliable database of the thermophysical and chemical properties of the fuel in its virgin as well as high burnup states. The wealth of information meanwhile noted from the irradiation studies of thoria-based fuels in the experimental reactors [8] will certainly provide the verification points of the simulation results. An impor­tant aspect in the simulation analysis will be the evaluation of the fuel-clad integrity or faultless containment of fuel inside clad. Such evaluation is quite established for the case of the conventionally used urania fuel, but it is not that much as will be called for the commercial implementation of the thoria-based fuels. The input of reliable physical and chemical information of thoria-based fuels in the performance analysis will strengthen the predictability of fuel behavior under the normal course of years’ long burning process inside reactor and also under off-normal situations like fuel containment problem due to clad failure out of stress corrosion cracking or loss of coolant accident. In fact, the physico­chemical database of fuel and fission products, and clad are frequently referred while planning the whole fuel cycle program from fuel design and fabrication and reprocessing of irradiated fuel to the management of nuclear waste. For countries like India that has meager resources of natural uranium, the realization of the whole fuel cycle for thoria-based fuels is necessary in order to make use of the fissile isotope U bred inside U/ Pu fuelled reactors.

The chemistry of the fission products (fps) is principally governed by the matrix within which these are produced. For understanding their chemistry in thoria fuel the available information on the chemical states of fps and their distribution inside urania fuel matrix are useful. The same set of fps with similar yields are formed and settle down inside the two fluorite lattices MO2 (M = Th+4,U+4) with similar crystal radii of the actinide cations. The yields of the fps for different fuels as given in Table 1 of “Thermochemistry of Thoria Based Fuel and Fission Products Interactions” subscribe to the general similarity in the two cases with the excep­tions that the thoria-based fuel results in comparatively more gaseous and less metallic fps. Nevertheless, there are some distinctive features in thoria. Chemi­cally, the distinctiveness originates from the rigidly four valency of Th in its compounds in condensed phases. This contrasts with U which is known to acquire higher valencies (four to six) in its oxides and compounds with alkali and alkaline earth fission products.

With increasing the oxygen partial pressure, urania undergoes oxidation from the stoichiometric UO2 to the hyperstoichiometric composition UO2+x whereas this aspect is absent in thoria. The valency rigidity of thorium results in increasingly less buffering of fission released oxygen and hence development of higher oxygen pressure in the thoria rich (Th, U)O2 fuels during their burnup. For the same reason the oxygen transport in thoria rich matrix is expected to be predominantly by self diffusion unlike the case in urania where the oxygen makes much faster transport through the chemical affinity driven diffusion process [11] to reach clad like zircaloy for its oxidation. The local regulation of fission generated oxygen owing to the impeded transport and poor buffering action in the fuel matrix practically rests on oxidation of the fps.

As against the above-mentioned undesirable features of furthering the fps’ oxidations in the irradiated fuel matrix, the thermophysical properties of thoria are superior in many respects to urania. The thermodynamic and kinetic analyses frequently refer to steady state as well as transient thermal profiles in the fuel pin. Thermal conductivity as a function of temperature and fuel composition is an important property in the analyses as it helps in establishing thermal profile at a given power rating. The thermal diffusivity (к = k/qCv) derivable from the con­ductivity (k), heat capacity (Cv), and density (q) is useful in calculating the relaxation time of thermal transients in power ramp when the thermal profile shoots up for a while resulting in augmented thermal stress in the fuel pin and promoting rapid redistribution and release of the gas and volatile fps. The thermal expansion properties help in understanding the fuel dilation relative to the clad and also in the analysis of fuel integrity in presence of thermal stress. The thermal stress developed over temperature differential AT is expressed by EaAT [12], where E is modulus of elasticity.

As compared to urania, the fuel dilates less and conducts more heat under a given temperature gradient. On the basis of available data of thermophysical properties of pure thoria and urania a comparative representation of the two material properties is included in Fig. 1 [13]. A look in Fig. 1 indicates that the thermal diffusivity (к = k/qCv) of thoria is even higher than urania so that under power ramp the thermal relaxation will be faster in thoria. Dutta et al. [13] has further evaluated thermal profiles of thoria and urania fuel pellets using the Code FAIR-TFC and one of their results is included in Fig. 2. These add to the merits of the thoria-based fuel. On the transport properties of oxygen, and gaseous and volatile fission products in urania and thoria matrices the reported information suggest subtle difference. The distinction in oxygen transport in the two oxide matrices has been indicated already. The combined involvement of vacancy and interstitial in the self diffusion of oxygen is reflected in the reported activation energy for O atom in the two oxide matrices; the energy barrier is higher in thoria (2.8 eV) than in urania (2.6 eV) [11, 14]. The reported value of anion interstitial migration energy (Qj) is significantly higher in thoria (3.27 eV) than in urania (2.6 eV) and vacancy migration energy (Qv) is comparable (*0.8-1.0 eV) [11, 14]. The gaseous and volatile species are expected to diffuse using interstitial and vacancy sites in the lattice. Th remaining strictly tetravalent in its oxide, the electronegative species such as I and Te show distinction in their diffusion behaviors in the two oxide matrices. In urania the diffusion is significantly influ­enced by O/M ratio; the oxygen hyper-stoichiometry augments the diffusion. Thermophysical properties of relevance to the evaluation of performance of thoria as fuel matrix are included in Table 1.

As for the fabrication, reprocessing, and waste management aspects of the thoria-based fuels there are again subtle distinction with urania. The fluorite phase of ThO2 in highly sintered state is chemically inert and this pose problem in acid

4 6

Pellet radius in mm

dissolution of irradiated fuel in its reprocessing. For improving the dissolution behavior defects are introduced in the fluorite lattice by doping with aliovalent oxides like magnesia, niobium oxide. The doping aids the sintering property also. The highly sintered state can be achieved then at significantly lower temperatures

Lattice parameter Theoretical density Linear thermal expansion

Thermal conductivity

Zero pressure bulk modulus at 298 K, and its pressure coefficient Bulk modulus at different porosity (volume) fractions (fp, 0.06-0.4) and temperatures (T, 298-1300 K)

Tensile strength at 298 К

Shear modulus 96.9 (1-2.12 fp) GPa at 298 К [17, 18], fp is porosity fraction

Standard enthalpy of formation at 298.15 К Standard entropy at 298.15 К Standard heat capacity

Melting point, heat of fusion

Sublimation paths vapor pressures of the sublimates (2400-2800 K)

Self diffusion coefficient of oxygen Self diffusion coefficient of Th/U Defects migration energies
559.730(3) pm [15]

9.9994 x 103 kg m~3

AL/L0 = -0.179 + 5.097 x 10~4(T/K) + 3.732 x 10~7(Т/КГ -7.594 x 10_11(T/K)3 [15]

1/(A + ВТ), A = 4.20 x 10~4 mKW-1, В = 2.25 x 10~4 mW4 [15] 196 GPa, and 5 respectively [16-19]

196 (1-2.21 fp) GPa [17]

196 [1.0230-14.05 x 10~5T Exp(-181/T)] GPa [17]

0. 082-0.102 GPa [17]

Rupture Modulus (MPa) = 440.963 cP°’3578 Exp(4.0858 fp) [17], d = mean dia. of grains in microns, fp = porosity fraction -1226.4 ± 3.5 kJ тоГ1 [15]

65.23 ± 0.20 J К_1тоГ1 [15]

C° = 55.962 + 51.2579 x 10~3 T-36.8022 x 10“5 T3 + 9.2245 x 10~9 T3-5.74031 x 105/T3 J КГ1 тоГ1 (298 < T < 3500 K), [13] and Cp(melt) = 61.76 J K_1 moP’festimated value) [15]

3651 ± 17 К [15], 90 kJ тоГ1 [15]

ThOo(s) = ThO(g) + O(g), ThOo(s) = Th02(g) log (PTho/Pa) = —36860/T(K) +13.15 [15] log (po/Pa) = —36800/T(K) +12.56 [15] log (PThor/Pa) = —35070/T(K) + 12.96 [13] log (D/пг s^1) = — 14362/T-3.35 [11] log (D/пг s^1) = —32715/T(K) -4.30 [11]

Anionic Vacancy [14] (0.78 eV) Cationic vacancy (7.04 eV) [14]

Interstitial [14] (3.27 eV)

Intertitialcy [14] (0.92 eV)

(*1873 K) and shorter period (*5 h) of thermal programming. Additional research inputs are involved in the fabrication, and reprocessing of thoria-based fuels. Similar inputs are also involved for incorporation of thoria containing nuclear waste inside the glass matrix for immobilization.

This book will cover the essential information on thermophysical, thermody­namic, and transport properties of oxide fuels with particular reference to thoria — based fuels. Besides it will cover the front and back end operations such as the procedures of fuel fabrication and characterization, and reprocessing and waste management of the reactor irradiated fuels. The thermophysical information includes the thermal conductivity/diffusivity and thermal expansion properties, heat capacity, and phase stability of the oxide fuels, oxygen potentials of the fuels, transport property of fission generated oxygen for understanding the possibility of its redistribution in fuel pin and uptake by clad, chemical states of the fission products and their distributions inside the pin, transport and release properties of the fission products xenon, and corrosive volatiles like iodine and tellurium. On fuel fabrication the book will cover fabrication procedures for different fuels such as ThO2, (Th, U)O2, (Th, Pu)O2, and also for non-oxide fuels for the sake of comparison. Different fabrication routes such as the conventional as well as modified powder-pellet route, sol-gel microsphere pelletization (SGMP) route, pellet impregnation route, and coated agglomerate-particle (CAP) route will be described. The procedures for fuel pin fabrication, compositional, and micro­structure characterization of the fabricated pellet/particulate, and also the safety aspects of handling thoria-based fuels will be outlined. On fuel reprocessing, the book will cover radiological problems encountered in irradiated thorium fuel reprocessing to recover 233U/Th and the details of Thorex process steps comprising of fuel decladding, dissolution, solvent extraction by TBP to recover 233U alone or both U and thorium, final purification of U product and its conversion to oxide. Variations and options in Thorex process to meet the different objectives and the possible areas of improvements in Thorex process and forthcoming develop­ments will be included. The last chapter of the book will present an overview of various types of waste streams, namely, low, intermediate, and high-level radio­active solid and liquid wastes that are generated in the spent fuel reprocessing. Various techniques used in the treatment of the radioactive wastes and safe disposals procedures of the treated wastes adopting different strategies for the LLW/ILW/HLW will be described. Elaboration will be made on the development and characterization of glass matrices used for immobilization of radionuclides present in the high-level waste. Aspects in the development of barium-borosilicate matrix for handling large concentration of sodium, iron, and sometimes sulfate present in the waste will also be covered.

The four chapters that immediately follows deal with the essential features of thermophysical, thermodynamic, and transport properties of thoria-based fuels, while the last three chapters provide the essentials of fabrication, reprocessing, and waste management of the fuel. In all the descriptions special reference has been made with the relevant features of the conventional urania/urania-plutonia fuels in order to bring home the merits and demerits of the thoria-based fuels.