Category Archives: Thoria-based Nuclear Fuels

Studies on ThO2

In India, ThO2 has been extensively used in PHWRs for neutron flux flattening of the initial core during start-up. During the last few years, ThO2 bundles have been utilized for this purpose in seven units of PHWR 220 including the two units each at Kakrapar Atomic Power Station (KAPS 1&2), Kaiga Atomic Power Station (KGS 1&2), Rajasthan Atomic Power Station (RAPS 3&4), and in RAPS 2 after mass coolant channel replacement. So far, some 232 thoria bundles have been successfully irradiated in the operating PHWR up to a maximum power of 408 kW and burnup of 13,000 MWd/Te HM without any failure [134]. Details of irradiation of ThO2 bundles in India are shown in Table 14. In-pile irradiation of Zircaloy-clad (Th, Pu)O2 fuel pins have been successfully carried out in the pressurized water loop of CIRUS research reactor. A six-pin cluster of free­standing Zircaloy-2 cladded ThO2-4 % PuO2 was successfully irradiated up to a burnup of 18,400 MWd/Te. Subsequently, two additional six pin clusters of col­lapsible Zircaloy-2 cladded pins containing high density ThO2 and ThO2-6.75 % PuO2 were successfully irradiated up to a burnup of 10,300 MWd/Te without failure. The peak pin-power rating was 40 kW/m [135]. Details of ThO2-PuO2 pin irradiation are shown in Table 15.

Although several isolated irradiation experiments on ThO2 and ThO2-UO2 materials have been conducted, the bulk of the irradiation data generated has come US from four goal-oriented programs [8, 12, 39, 136]:

1. Boiling water reactor (BWR) Program, which culminated in the irradiations performed in the BORAX-IV and Elk River reactors.

2. Thorium Utilization Program, which included the irradiation of vibratory compacted and pelletized fuels.

Table 14 Irradiation of ThO2 bundles in Indian PHWRs


Number of bundles

















Table 15 Details of the irradiation of ThO2-PuO2 fuels


Cladding type

Burnup (GWD/t)

Linear rating (kW/m)

ThO2-4 % PuO2




ThO2-6.75 % PuO2




3. Babcock and Wilcox developmental work on pressurized water reactor (PWR) fuels.

4. Light water breeder reactor (LWBR) Program at Bettis Atomic Laboratory which resulted in the core loading of the Shippingport Reactor.

The majority of experiments were performed on fuels with less than 10 wt% UO2. From the above experiments and other experiments carried out in India, Russia, Canada, Japan, and Korea the following conclusions can be drawn about thermophysical properties.

Thoria-based Nuclear Fuels

It is now amply clear that nuclear energy can be a sustainable source of clean energy for the whole world if we take into account both the fissile and the fertile material resources availability. If, however, uranium-235 is taken as the only fissionable nuclide from which we can generate nuclear power, the exploitable resources are not adequate to supply fuel for the growing demand of the nuclear power industry beyond a century. The adoption of closed fuel cycle to utilize plutonium-239 which is generated as a transmutation product of uranium-238 in the spent nuclear fuel is, therefore, essential in multiplying the fuel resource base by over 50 times. Thorium-232, the other naturally occurring fertile material can enhance the fissile resources still further several fold. In addition, the use of uranium-233, the fissile product of thorium, offers several advantages in reducing radiotoxic waste burden and in providing a proliferation resistant fuel. Worldwide R&D efforts in establishing the reactor technology with thorium fuel cycle have significantly grown in the recent past mainly because of these factors. India being endowed with a huge reserve of thorium minerals has given the due emphasis on research on thorium fuel cycle and has drawn the roadmap of thorium utilization in her long-term plan for the growth of nuclear energy.

This book gives a comprehensive treatment of various aspects of thoria-based fuels and both the front and the back end of the fuel cycle associated with it. Each chapter is written by experts who have many years of experience in their respective fields. At the same time, the continuity of the theme and coherence among the chapters are maintained. The editors must be complimented for structuring the book very well and make it suitable not only for those who are entering into research in the nuclear fuel cycle area but also for practicing sci­entists. The authors have paid attention to technological details while presenting the scientific basis from a fundamental standpoint. The fact that the authors have direct research and operating experience in the areas which are covered in their respective chapters gets reflected in the clarity of their presentations. There are only a few in the world who have delved into thorium fuel cycle, fabrication of thoria based fuel, reprocessing of high burn up fuels and their waste management
issues. As they have joined hands to prepare this book entitled Thoria-based Nuclear Fuels: Thermophysical and Thermodynamic Properties, Fabrication, Reprocessing and Waste Management, the product that has emerged is not only unique but also extremely valuable.

image2Mumbai, India, April 2013

Srikumar Banerjee Homi Bhabha Chair Professor Bhabha Atomic Research Centre (Former Chairman, Atomic Energy Commission, & Secretary to the Department of Atomic Energy,

Government of India)


The thermal linear expansion of ThO2 is well established. Touloukian et al. [61] list more than 34 different experimental determinations which are in excellent agreement and recommended the following equation (150-2,000 K):

% AL/Lo = -0.179 + 5.097 x 10-4(Г/K) + 3.732 x 10-7(Г/K)2-7.594

x 10-11(T/K)3; (25)

Touloukian states that the equation has an accuracy of ±3 % or less. Hoch and Momin [72] showed that their data, obtained by XRD measurements of thermal expansion of ThO2, is in very good agreement with that obtained by Ohnysty and

Temperature (K)


Expansion coefficient, x10 6 K 1































































Rose [73] by dilatometry. Their data also agreed well with that of Aronson et al. [74] and Hirata et al. [75]. Hoch and Momin [72] recommended the following equation for ThO2 (293-2,373 K):

% AL/L0 = -0.2426 + 7.837 x 10-4(Г/K) + 9.995 x 10^8(Г/K)2 (26)

The results obtained from the above equations are almost identical. Hoch and Momin concluded from their studies that the lattice defects in ThO2 were probably of the Frenkel type.

The extrapolation data to higher temperatures other than those covered by measurements need caution [76]. The equation of Touloukian is valid in the range of 150-2,000 K and that of Hoch and Momin [72] should be used for extrapola­tion. The recommended values of thermal expansion of are shown in Table 10. Thermal expansion curve for ThO2 is shown in Fig. 7.


In nuclear power generation establishing the performance of a fuel calls for collective effort of scientists and technologists for addressing various issues of front and back ends of the fuel cycle. These issues pertain to all the operational aspects for smooth functioning right from fuel resourcing, mining and purification, fuel pin fabrication, reactor irradiation, fuel reprocessing, and management of radioactive wastes. Safety, security, and economy are to be met in the fuel cycle management. Typically nuclear fuels are irradiated in reactors for several decades for deriving power from them. Therefore, a thorough understanding of thermal and thermodynamic properties of the fuel materials is necessary for the evaluation of its performance under these conditions. Reliable data acquisitions are thus carried out for the thermal expansion, thermal conductivity, heat capacity, phase stability of the fuel, and thermodynamic and transport properties of fission products that are formed in the fuel matrix. With the additional knowledge of the transport properties of fission-released oxygen in the case of oxide fuels, the chemical states of the redistributed fission products are evaluated for ensuring containment of the cladded fuel.

Many countries, particularly those having abundant resource of thorium, envisage nuclear power production from this fertile-actinide. The envisaged policy for the thorium-based reactor technology is toward reducing burden to the enriched uranium-based conventional fuels. Thorium-based fuels have the general merits of greater abundance of the element on earth crust, superior physical and nuclear properties, particularly in oxide form, better resistance to weapons proliferation and lesser production of heavier actinides in reactor irradiation. There has been worldwide effort to establish the thorium-based reactor technology and thus a large database on the thermophysical, thermodynamic and transport properties of thoria based fuels, and detailed technical information on the fuel fabrication, repro­cessing and waste management exists in the literature. In this context, it is nec­essary to consolidate the accumulated information at one place, in the form of a book that essentially covers the scientific and technological information on all the stated aspects of thoria-based fuels. With this objective, the chapters in this book are organized accordingly and are written by experts in the respective fields, who collectively contributed to the thorium utilization program in India. The arrangement of chapters has been carefully planned to provide the readers with adequate state-of-art knowledge regarding thoria fuels. We believe that the

scientific and technical information in this book will serve as a ready reference to researchers and technologists working in the field of thorium utilization.

Editors of this book sincerely acknowledge the authorities of the Department of Atomic Energy, Government of India, who have appreciated and approved the idea of making such publication that presents a consolidated view of R&D on the thorium-based fuel cycle. We would like to express our gratitude to all the con­tributors and to the staff of Springer-Verlag for their generous assistance without which this book could not have been successfully published.

Mumbai, India, April 2013 D. Das

S. R. Bharadwaj


Substitution of uranium for thorium in ThO2-UO2 system results in the decrease of lattice parameter, melting point, oxygen potential, thermal conductivity, while the density and the linear thermal expansion show an increase. In the uranium con­centration range y > 0.2, less data are available but the same trend is expected. It was mentioned earlier that, at room temperature, the lattice parameter decreases linearly from 100 % ThO2 to 100 % UO2. Based on the results of calorimetric studies, thermal expansion studies, and other thermodynamic measurements on urania-thoria solid solutions [40], it is suggested that stoichiometric urania-thoria solid solutions are nearly ideal at least up to 2,000 K. Therefore, one can assume that this linear decrease in the lattice parameter also exists at high temperature. This linear decrease can only exist when the linear thermal expansion of (Th1-yUy)O2 (0 < y < 1) equals the linearly interpolated value of that of ThO2 and that of UO2.

The thermal linear expansion of the (Th1-yUy)O2 solid solution has been studied in much less detail than that of the pure compounds. Konings et al. [39] reviewed the thermophysical properties of ThO2-based fuels. They concluded from the results of eight studies on (Th1-yUy)O2, and suggested the equation of the type:

AL/L0 = (8.1635 x 10-4 + 3.8325 x 10-4y + 5.2423 x 10-4y2) (T/K — 298.15)

+ (1.2144 x 10-7 + 1.4936 x 10-8y + 1.5633 x 10-7y2)(T/K — 298.15)2,


where, the thermal linear expansion AL/L0 in %, and y is the molar fraction of UO2.

Bakker et al. [38] have recommended the percentage linear thermal expansion data of (Th1-yUy)O2 (0 < y < 1) by obtaining the linear interpolation of the values of Touloukian [61] and Martin [47] and obtained the following relations in two different set of temperature ranges:

Подпись: •T (30) (dL/L0) x 100 = -0.179 — y 0.087 + (5.097 x 10-4 + y4.705 x 10-4)

+ (3.732 x 10-7 — y 4.002 x 10-7) T2 — (7.594 x 10-11 — y 11.98 x 10-11) • Г3

(for 273 K < T < 923 K),

(AL/L0) x 100 = -0.179 —y 0.149 + (5.097 x 10-4 + y 6.693 x 10-4) • T

+ (3.732 x 10-7 — y 6.161 x 10-7) Г2 (31)

— (7.594 x 10-11 — y 19.784 x 10-11) • T3

(for 923 K < T < 2,000 K).

Momin et al. [60] measured lattice thermal expansion of (Th, U)O2 system by X-ray diffraction method. They obtained coefficient of expansion data for pure ThO2 and (Th0.8U0.2)O2 to be 9.5 x 10-6 K-1 and 7.1 x 10-6 K-1, respectively, in the temperature range 298-1,600 K. It was observed that the coefficient of thermal expansion of (Th08U02)O2 is lower than either of ThO2 and UO2, which is quite unreasonable.

Tyagi et al. [77] found CTE values for ThO2 and ThO2-2 wt% UO2 to be 9.58 x 10-6 and 9.74 x 10-6 K-1, respectively, in the temperature range of 298-1,473 K. The CTE value reported in IAEA-TECDOC [40] for ThO2 in the temperature range 300-1,473 Kis9.732 x 10-6 K-1 andforThO2-4 wt%UO2itis 9.85 x 10-6 K-1; both these values were found to be in close agreement with those reported by Tyagi et al. The CTE value 10.33 x 10-6 K-1 for composition (Th087U0.13)O2 in the temperature range 298-1,973 K as reported by Anthonysamy et al. [78] matches well with the value obtained in the IAEA study for ThO2-10 % UO2 which was found to be 10.21 x 10-6 K-1 in the temperature range 300-1,773 K. The average linear thermal expansion coefficients for (Th045U055)O2 and (Th0 09U0 9i)O2 were measured to be 10.83 x 10-6 K-1 and 11.45 x 10-6′ K-1, respectively, in the temperature range between 298 and 1,973 K. These data clearly show that thermal expansion coefficients increases with increase in UO2 content in ThO2-UO2 system. Figure 9 shows % thermal expansion plot of some typical ThO2-UO2 solid solutions.

Подпись: Fig. 9 Percent thermal expansion plot of some typical ThO2-UO2 solid solutions [78]. (permission from Elsevier)

The coefficient of expansion data of Momin et al. [60], Springer et al. [79], Turner and Smith [80], Kempter and Elliot [56] and Lynch and Beals [81] show a wide scatter of data points when plotted against composition. Rodriguez and Sundaram [82] in their review article reported an average linear thermal expansion coefficient of 9.67 x 10-6 K-1 for ThO2 (293-2,273 K) and 12.5 x 10-6 K-1 for (Th08U0.2)O2 (1,100-2,400 K). Powers and Shapiro [83] reported the same average linear thermal expansion coefficient value of 9 x

UO2 and (U0.064Th0936)O2. They obtained lower coefficient value (8 x 10-6 K-1 up to 1,073 K) for (Th0 8U0.2)O2.

Kutty et al. [84] measured thermal expansion of ThO2, ThO2-4 % UO2, and ThO2-20 % UO2 pellets fabricated by (Coated Agglomerate Pelletization) CAP route using ThO2 and U3O8 powders as the starting materials. They reported that the thermal expansion of ThO2-20 % UO2 pellet was different from that of ThO2 and ThO2-4 % UO2, e. g., it increased more rapidly with increasing temperature in the temperature range of 1,000-1,500 C which they attributed to the loss of oxygen of (Th, U)O2+x above 1,000 C. The thermal expansion behavior of polycrystalline samples of ThO2-3.45 % UO2 and SIMFUEL corresponding to the burnup of 43,000 MWd/Te has been investigated from room temperature to

I, 473 K, and for SIMFUEL corresponding to burnup of 28,000 MWd/Te has been investigated from room temperature to 1,173 K, using a high-temperature X-ray diffraction (HTXRD) by Bhagat et al. [85]. They reported that SIMFUEL has higher thermal expansion than ThO2-3.45 % UO2 and this is related to the higher thermal expansion coefficient of dissolved rare earth oxides and also to the lower melting point of SIMFUEL matrix.

The mean linear thermal expansivity for ThO2-SmOi.5 solid solutions con­taining 17.9, 41.7 and 52.01 % of SmOi.5 were determined by Subramanian et al. [86] in the temperature range 298-2,000 K. The mean linear thermal expansion coefficients for ThO2-SmO15 solid solution were found to be 10.47, 11.16, and

II. 45 x 10-6 K-1, respectively. The synthesis, characterization, and lattice thermal expansion studies of the ThO2-Nd2O3 phase with general compositions Thi-xNdxO2 — x/2 are reported by Mathews et al. [87]. The lattice thermal expan­sion (293-1,473 K) behavior of the solid solutions has been investigated by high temperature XRD and found to show a gradual increase with increasing content of NdOi.5 in Thi-xNdxO2-x/2 series. The lattice thermal expansion behavior of a number of single-phase compositions of CeO2-ThO2-ZrO2 in the temperature range from 293 to 1,473 K, as investigated by high-temperature XRD are reported by Grover et al. [88]. The average lattice thermal expansion coefficient of pure thoria was found to be 9.58 x 10-6 K-1, which increased to 11.91 x 10-6 K-1 in the compositionTh0.05Ce0.90Zr0.05O2.

Momin et al. [60] studied thermal expansion behavior of ThO2 and (Th08U02)O2 with 20 wt% Ln2O3. Ln2O3 contained oxides of La, Nd, Ce, Y, Sm, Gd, and Eu in equal proportions. Authors found that average thermal expansion coefficient of the solid solutions of ThO2 and (Th0.8U0.2)O2 with 20 wt% Ln2O3 show an increase as compared to those of the parent compounds. Authors related the higher values of coefficient of expansion to the partial substitution of U4+ or Th4+ with Ln3+ resulting in weakening the interatomic bonding in the solid solution matrix. Grover et al. [88] found that coefficient of linear thermal expansion of (Th, Ce, Zr)O2 is higher than ThO2 and increases with increase of cerium and zirconium content in (Th, Ce, Zr)O2. Dilatometric measurement on ThO2-10.09 % UO2 and ThO2-20.02 mol% UO2 by Springer et al. [79] and XRD determination on ThO2-50.05 % UO2 by Kempter and Elliot [56] have been reported. Variation in expansion with compo­sition in any case is reported to be quite small.


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


(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


Подпись: Fig. 2 Thermal profiles in thoria and urania based fuels image3,image4,image5

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

Подпись: Crystal structure Fluorite (CaF2 type) structure, space group Fm3m (No. 225) [15]
Подпись: Introduction

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.


ThO2 and PuO2 form a continuous series of solid solutions over the entire range of composition. At the Pu-rich end, mixed oxide may be heterogeneous if prepared under reducing conditions, as a result of the formation of Pu2O3. The lattice parameter of (Th1-yPuy)O2 decreases linearly from pure ThO2 to pure PuO2 [26]. Lattice parameters of (Th, Pu)O2 with various PuO2 contents are given in Table 11. Assuming ideal solid solution behavior at high temperatures for ThO2 and PuO2, it would be expected that this linear decrease in lattice parameter would also happen at elevated temperatures.

The available data on ThO2-PuO2 are scanty. One way to overcome this problem is to use CeO2 in place of PuO2 as they both have quite similar physico­chemical properties viz., ionic radii in octahedral and cubic coordination, melting points, standard enthalpy of formation and specific heat etc. Thus, the plutonium chemistry can be well simulated using CeO2 in place of highly active PuO2. Mathews et al. [89] have recently measured bulk thermal expansion of (Th, Ce)O2 system. Bulk and lattice thermal expansion studies on (Th1-yCey)O2 (y = 0.0, 0.04, 0.08 and 1.0) were carried out by dilatometry and high temperature XRD from room temperature to 1,123 and to 1,473 K, respectively. The average linear thermal expansion coefficients of ThO2, Th096Ce004O2, Th092Ce008O2, and CeO2 were found to be 9.04 x 10-6, 9.35 x 10-6, 9.49 x 10-6, and 11.58 x 10-6 K-1, respectively, between 293 and 1,123 K. Some data on (Th1-yPuy)O2 generated at BARC was reviewed in IAEA-TECDOC [40]. Thermal expansion curve for (Th1-yPuy)O2 for y = 0.02, 0.04, 0.06, 0.10 are shown in Fig. 10.

The thermal expansion of the solid solutions (Th1-yPuy)O2 could be reasonably approximated at various temperatures by taking linear interpolated expansion data of ThO2 and PuO2 as per their weight fraction. IAEA-TECDOC [40] used “interpolation method,’’ using the recommended equation by Touloukian [61] for ThO2 and the following equation for pure PuO2 as recommended by MATPRO. The equation for linear strain calculations is as given below:

s = K • T — K2 + K • exp(-Eo/ksT); (32)

PuO2 (mol%)

Lattice parameter (pm)

































Table 11 Lattice parameter of ThO2-PuO2 solid solution at 298 K [26]

Подпись: Fig. 10 Thermal expansion curves for (Th1-yPuy)O2 for y = 0.02, 0.04, 0.06 and 0.10 [40]. (permission from IAEA)

where, e is the linear strain which is taken as zero at 300 K, T is the temperature (K), kB is Boltzman’s constant (1.38 x 10-23 J/K), and ED, K1, K2, and K3 are constants having values 7 x 10-20 (J), 9 x 10-6 (K-1), 2.7 x 10-3 (unit less), and 7 x 10-2 (unit less), respectively.

Percentage linear thermal expansion for (Th1-yPuy)O2 (0 < y< 1) obtained by linear interpolation of the data of ThO2 and data for PuO2 and can be expressed as (in the temperature range of 300-1,773 K) [40].

(Th1-yPuy)O2 where 0 < y <1:

(AL/L0) x 100 = — 0.179 — 0.049 • y + (5.079 x 10-4 + 2.251 x 10-4 • y) • T + (3.732 x 10-7 — 2.506 x 10-7 • y) • T2 + (-7.594 x 10-11 + 12.454 x 10-11 • y) • T3


As part of thorium-based fuel development program for fast breeder reactors, the thermophysical properties of mixed thorium-plutonium oxide pellets of both thorium — and plutonium-rich compositions were evaluated in India [58]. The plutonium-rich mixed oxide pellets contained 70-80 % PuO2 which could be considered as candidate fuel for small LMFBR core like the operating fast breeder test reactor (FBTR). The thorium-rich compositions contained 20-30 % PuO2 which could be considered as alternative fuel for large LMFBRs like the forth­coming prototype fast breeder reactor (PFBR-500). The mixed oxide pellets were prepared by ‘‘powder-pellet’’ route involving mechanical mixing of ThO2 and PuO2 powders followed by cold pelletization and high temperature sintering. Small amount of Nb2O5 (0.25 wt%) or CaO (0.5 wt%) powder were used as ‘‘sintering aid’’ and admixed with the powder during co-milling. The coefficient of thermal expansion of mixed (Th03Pu07)O2, (Th07Pu03)O2, and (Th08Pu02)O2

Fig. 11 Thermal expansion 14

Подпись:curves for high Pu bearing (Th1-yPuy)O2 samples [58].

(permission from IAEA) 12





Ш 8 —


6 4

were evaluated by a high-temperature dilatometer and is summarized in Fig. 11. XRD pattern of ThO2 and the pellets containing lower amounts of PuO2 (30 % PuO2) sintered in either Ar or Ar-8 % H2 showed only single-phase isostructural with fluorite phase. But ThO2-PuO2 pellets with higher plutonium content such as ThO2-50 % PuO2 and ThO2-75 % PuO2 pellets sintered in either Ar or Ar-8 % H2 showed the presence of two phases. In addition to the phase that is isostructural with PuO2 (fluorite), another phase which is isostructural with bcc a-Pu2O3 has been observed [90]. Hence, no conclusion could be drawn from the above results.

Thermophysical Properties of Thoria-based Fuels

T. R. Govindan Kutty, Joydipta Banerjee and Arun Kumar

Abstract The behavior of nuclear fuel during irradiation is largely dependent on its thermophysical properties and their change with temperature and burnup. Experi­mental data on out-of-pile properties such as melting point, density, thermal con­ductivity, and thermal expansion are required for fuel design, performance modeling, and safety analysis. The variables that influence the out-of-pile properties are fuel composition, temperature, porosity, microstructure, and burnup. Among the above-mentioned properties, thermal conductivity of nuclear fuel is the most important property which influences almost all the processes such as swelling, grain growth, and fission gas release, and limits the linear power. The changes in thermal conductivity occur during irradiation by the formation of fission gas bubbles, porosities, build-up of fission products, and by the change of fuel stoichiometry. Melting point plays a crucial role in determining the power to melt the fuel and decides the operating linear heat rating. The coefficient of thermal expansion (CTE) is needed to calculate stresses occurring in the fuel and cladding on change in temperature. In safety analysis, the values of thermal expansion data are required in determining the gap conductance and the stored energy.

1 Introduction

The behavior of nuclear fuel during irradiation is largely dependent on its phys­icochemical properties and their change with temperature and burnup [1]. Thermal conductivity is an important parameter to understand the performance of the fuel

T. R. Govindan Kutty (H)

Formerly at Radiometallurgy Division, Bhabha Atomic Research Centre,

Mumbai 400085, India

e-mail: trgovindankutty@gmail. com

J. Banerjee • Arun Kumar

Radiometallurgy Division, Nuclear Fuels Group, Bhabha Atomic Research Centre, Mumbai 400085, India

D. Das and S. R. Bharadwaj (eds.), Thoria-based Nuclear Fuels, Green Energy and Technology, DOI: 10.1007/978-1-4471-5589-8_2, © Springer-Verlag London 2013

pins under irradiation [2]. It is highly dependent on physical structure, state, chemical composition, and is one of the most important properties for predicting fuel and material performance [3]. If the thermal conductivity is low, the tem­perature gradient in the radial direction of the fuel pellet is large which results in high temperature at the central part of the fuel pin [2, 3]. The thermal conductivity of nuclear fuel influences almost all important processes such as fission gas release, swelling, grain growth etc. and limits the linear power [4, 5]. The changes in thermal conductivity occur during irradiation by the formation of fission gas bubbles, build-up of fission products, and by the change of oxygen-to-metal ratio (O/M) [6]. Hence, the knowledge of thermal conductivity is needed to evaluate the performance of nuclear fuels. The coefficient of thermal expansion (CTE) values is needed to calculate stresses occurring in the fuel and cladding. If the thermal expansion varies considerably between the fuel and cladding, then stresses will be accumulated during the thermal cycling [7]. This can lead to deformation of the cladding and eventually may result in the breakage of the cladding. Hence, precise evaluation of CTE data of the fuel is needed.

Other important thermophysical properties to be considered are melting point and density. Thorium and uranium oxide fuels used in nuclear reactors have very high melting points, but low density and they suffer from poor thermal conduc­tivity, because in these insulating oxides only phonons (lattice vibrations) conduct heat. Understanding the physics underlying transport phenomena due to electrons and lattice vibrations in actinide systems is an important step toward the design of better fuels [8].

Thermophysical properties of materials depend on various factors, such as microstructure, porosity and its distribution, thermal treatment employed, pro­duction technology used, radiation exposure undergone, and other unidentified factors leaving aside the temperature effect [1]. Improving the technology for nuclear reactors through better computer codes and more accurate data of materials property, which can contribute to improved performance as well as economics of future plants by getting rid of currently used large design margins. Accurate representations of thermophysical properties under relevant temperature and neu­tron fluence conditions are therefore, necessary for evaluating reactor performance under normal operation and accidental conditions [2].

Prior to deploying new fuels and structural materials in a nuclear reactor, its thermophysical properties must be known. The fuel temperature is determined by the thermal conductivity. Such properties of the fuel are not constant during the irradiation period in the reactor, but change with the burnup. Therefore, the evaluation of the thermophysical properties of fuel, including a reliable uncertainty assessment, is required by the nuclear reactor design [3]. A high confidence level on the fuel performance can only be reached from a good interpretation of the irradiation data followed by post-irradiation examinations. A prerequisite for this is to have data on out-of-pile properties such as thermal conductivity or thermal diffusion that allows to understand the influence of parameters such as tempera­ture, temperature gradient, stress, stress gradient, fission rate, and impurities that are effective during the operation. Safety analyses are required by regulatory authorities to prove that the fuel can be burned safely in the reactors. These safety analyses require calculations with safety codes that need the appropriate thermo­physical properties of the fuel. These important informations are used by ther­mohydraulic codes to define operational aspects and to assure the safety, when analyzing various potential accidental scenarios. For each property, the variables that influence the property are to be described, followed by a review of the available data and correlations. Variables considered are temperature, composi­tion, porosity (p), burnup (B), and oxygen-to-metal (O/M) ratio [9, 10].

Due to its higher thermal conductivity, during normal operation, ThO2-based fuel will operate with somewhat lower fuel temperatures and release less fission gas than UO2 fuel at corresponding powers and burnups. This will allow for higher prepressurization and thereby minimizes cladding creep down and fuel cladding mechanical interactions at high burnup and thereby possibly allow for higher burnup use of this material. During an accident such as a large loss-of-coolant accident (LOCA), ThO2-UO2 fuel will have less stored energy but a slightly higher internal heat generation rate than UO2 fuel at similar power levels [11]. As a result, certain parameters for accident evaluation such as the maximum cladding temperature and the timing of fuel rod rupture are expected to be slightly different [8]. These expected differences in behavior between ThO2-UO2 fuel rods and UO2 fuel rods need to be quantified for an objective evaluation of the performance of ThO2-UO2 fuel. The mixed ThO2-UO2 fuel reduces the amount of total plutonium production by a factor of 3.2 and the 239Pu production by a factor of about 4, when compared with conventional UO2 fuel irradiated to 45 GWD/t [9]. The plutonium that is produced in the mixed ThO2-UO2 fuel is high in 238Pu, producing copious amounts of decay heat and spontaneous neutrons making it proliferation resistant. A mixture of ThO2 and UO2 is much more resistant to long-term corrosion in air or oxygenated water than UO2. Thus, ThO2-UO2 is a superior waste form if the spent fuel is slated for direct disposal rather than reprocessing.

Thermal Conductivity

Подпись: where v is the linear rating. Подпись: V image17 Подпись: (34)

Among the various thermal properties, thermal conductivity is the most useful property for the nuclear scientist. It is the ability of the material to transfer heat from a region of high temperature to a region of low temperature. In normal conditions, thermal conductivity and linear power determine the peak fuel oper­ating temperature. Under the accident conditions, the thermal conductivity of the fuel determines the maximum permissible linear rating, vmax, if central melting is to be avoided [10, 38]. The thermal conductivity, k, allows the determination of centre temperature of fuel, Tc, when the surface temperature Ts, is known by using the conductivity integral,

For ceramic oxide systems, two contributions are used to describe the behavior of thermal conductivity with temperature: (i) the phonon-phonon interaction and (ii) the density of defects (phonon scattering centers) in the lattice [91]. For temperatures below 1,900 K, the contribution of the free electrons to the thermal conductivity can be neglected.

Melting Point

A very important thermophysical property to be considered for an engineering material, like nuclear fuel, is its melting point. The onset of melting at the cen­terline of the fuel rod has been widely accepted as an upper limit to the allowable thermal rating of nuclear fuel elements [11, 12]. The melting point must be taken into account when considering a new fuel, as it limits the power that can be extracted from the fuel element. The knowledge of the melting point is also important in the fabrication of chemically homogeneous pellets like thoria-urania since ThO2 (3,663 K) and UO2 (3,100 K) have high melting points and relatively low diffusion coefficients at normal sintering temperatures [13].

As a pursuit for the better fuel, it is crucial to understand the underlying transport phenomena due to electrons and lattice vibrations in actinide systems.

According to the Lindemann criterion, solids with large Debye frequencies have high melting points [14]. This is typically found in insulators where atomic bonds are strong due to lack of free electrons. Thorium contains no occupied 5f states while uranium has two unpaired 5f valence electrons, and therefore uranium and thorium possess very different electronic and chemical properties. Since melting point is an important property, it is worth considering from the standpoint of the bonding present in actinide elements [15]. The highest melting points for the actinide metals are for Th and Pa metals. The effect of f-orbitals on the melting point is maximized with Np and Pu; both have very low melting points, which are believed to be a reflection of f-orbital repulsion [16]. Uranium has multiple oxidation states (3+ through 6+) which allow UO2 to be easily oxidized to U3O8 or UO3, by incorporating interstitial O atoms. In contrast, thorium only exhibits one oxidation state (4+) and hence cannot be oxidized beyond ThO2 [9, 17]. Among the actinide oxides, only ThO2 is a white insulating solid and the other AnO2 solids are all dark and opaque and poorly conducting. While in UO2 the 5f electrons, which occupy an energy band from 1.37 to 1.50 eV cause a strong visible light absorption resulting in a dark gray color of this oxide. The consequent absence of low electrons in the valence band is the cause of the high transparency of stoichiometric thoria and the low spectral emissivity in the visible range at room temperature [17].

Thorium dioxide exists up to its melting point as a single cubic phase with the fluorite crystal structure, isomorphous, and completely miscible with UO2. Unlike UO2, ThO2 does not dissolve oxygen to a measurable extent. Therefore, it is stable at high temperature in oxidizing atmosphere. On prolonged heating to 1,800-1,900 °C in vacuum, it blackens with loss of oxygen, although the loss is insignificant to be reflected in lattice parameter or in chemical analysis. On reheating in air to 1,200-1,300 °C, the white color is restored.

The melting points of the nuclear fuels are shown to be influenced by the following factors: stoichiometry and composition, irradiation dose, impurities and their contents.

2.1 ThO2

The melting point of ThO2 was experimentally measured or estimated by several authors [18-26]. Their results are summarized in Table 1. As it can be seen from the Table, the reported values vary from 3,323 to 3,803 K. Peterson and Curtis [26], in their compilation of data on thorium-based ceramics, arrived at two dif­ferent values, e. g., 3,573 ± 100 K from the work of Lambertson et al. [21] on ThO2-UO2 system and 3,663 K from the work of Benz [22] on Th-ThO2 system. Lambertson et al. [21] first estimated the melting point of ThO2 to be between 3,558 and 3,828 K and subsequently arrived at an intermediate value of 3,623 K by extrapolating the melting point data of (Th, U)O2 compositions corresponding to zero UO2 content. They further refined their data by introducing some corrections

Table 1 Melting point of ThO2 determined by various authors



Melting point (K)


Ruff et al. [18]



Wartenberg and Reusch [19]



Geach and Harper [20]

3,323 ± 25


Lambertson et al. [21]

3,573 ± 100


Benz et al. [22]

3,663 ± 100


Christensen [27]



Chikalle et al. [23], TECDOC-1319



Rand [25]

3,643 ± 30


Belle and Berman [12]

3,640 ± 30


ITU [25]

3,640 ± 20


Ronchi and Hiernaut [24]

3,651 ± 17

for the liquidus/solidus curve to effect a curvature correction for the pure ThO2 end to that of pure UO2 end of the temperature—composition diagram. Their final recommended data was 3,575 K, which is in good agreement with the data 3,543 K recommended by Christensen [27] from his experimentally measured melting point data on ThO2-UO2 system and subsequent extrapolation to zero UO2 content. Rand [25], however, disagrees with the curvature corrections made by others on the thoria or urania rich side of the temperature composition curve. He justified that the curvature need not be same at both the terminal compositions and the difference could be due to the loss of ‘O’ from UO2 in urania-rich side, which is different for thoria-rich side. He recommended a value of 3,643 ± 30 K. Belle and Berman [12] used 3,640 K as the melting point of ThO2, recommended by Rand [25] in his work on ThO2. Ronchi and Hiernaut [24] had recently measured the melting temperature of ThO2 (both stoichiometric and hypostoichiometric) material experimentally by heating a spherical sample by four symmetrically spaced pulsed Nd YAG laser and observing the cooling/heating curve with time. For stoichiometric ThO2, the measured melting point was found to be

3.651 ± 17 K. The data of Ronchi and Hiernaut [24] reasonably agrees with the data generated by Benz [22] (3,660 ± 100 K) and is also close to that recom­mended by Rand [25] (3,643 ± 30 K). All these values are markedly different from those of Lambertson et al. [21]. It is also well understood that the curvature difference at the uranium — and thorium-rich side of the temperature versus com­position diagram is quite justifiable and was attributed to the loss of oxygen. Hence, the recommended melting temperature of ThO2 should be taken as

3.651 ± 17 K, and is in fairly good agreement with majority of the previous studies. The value measured at the Institute for Transuranium Elements (ITU) (3,640 ± 20) K, is very close to the value reported by Rand [25].

Measurements of the cooling curves of molten ThO2 and ThOi.98 reveal that the stoichiometric compound melts congruently at 3,651 K, while the hypostoichio — metric oxide displays a liquidus at 3,628 K and a solidus transition at 3,450 K. Ronchi et al. [24] conducted pulse-heating experiments on thoria and showed that this compound exhibits a premelting transition at 3,090 K whose features are analogous to those observed in other ionic compounds having fluorite type structures. A class of diatomic compounds which crystallize in the face-centered cubic fluorite lattice (space group Fm3m), and whose component atoms have very different mobilities, exhibit at a temperature corresponding to about 80 % of the absolute melting temperature, a premelting transition. The discovery of this transition in UO2 [12] gave rise to a number of investigations aimed at defining its nature and possible effects on the high temperature properties of this technologi­cally important material. In ThO2-x, the dependence of the transition temperature, Tc on stoichiometry was found to be weaker than hypostoichiometric urania and exhibits an opposite trend, with Tc decreasing with x.

To conclude, the melting temperature of ThO2 recommended from this assessment is 3,651 ± 17 K and is in fairly good agreement with majority data available in the literature.