Viability of Thorium-Based Reactors

Transcription

1 Viability of Thorium-Based Reactors by Atnatiwos Zeleke Meshesha A dissertation submitted to the Department of Physics, University of Surrey, in partial fulfilment of the degree of Master of Science in Radiation and Environmental Protection Supervisor: Dr. P. H. Regan Department of Physics Faculty of Engineering & Physical Sciences University of Surrey September 2007 Atnatiwos Zeleke Meshesha 2007

2 ABSTRACT Viability of Thorium Based Reactors It was at the earliest times of nuclear technology that interest in thorium was developed as a supplement to limited uranium reserves. Research and development activities have been conducted in a number of countries and many remarkable progresses resulted; prototype reactors such as high temperature gas cooled reactors (HTGR) and thermal breeding in light water breeder reactor (LWBR) has been demonstrated. Recently, new considerations have also revived interest in thorium fuels. These may be due to growing global energy demand, the issue of climate change and public concern to radiotoxic waste and proliferation to nuclear weapons. In addition, the burning of plutonium stockpiles and energy sustainability can also contribute. This review examines the viability of thorium-based reactors, discussing types and features of thorium fuel cycles and analyses their potentials and drawbacks. Interesting earlier thorium-based nuclear systems are recalled and advanced and innovative nuclear systems and fuel cycles, such as accelerator-driven system (ADS), high temperature reactors (HTR) and molten salt breeder reactor (MSBR) examined. It was found that thorium-based fuels have many advantages, such as in contributing burning of excess plutonium, generating less long-lived waste, permitting higher burn-ups to be achieved, demonstrating acceptable proliferation- resistance, permitting higher temperatures to be reached, and sustainability in view of limited uranium reserves. However, much improvement on identified challenges, such as its compatibility with existing nuclear systems and the long process of thorium spent fuel separation is still expected. Further research and development to justify its practicality and to test the anticipated safety and fuel performance is also required. Thorium fuel with its advantages and mainly due to its superior neutronic properties, have potential in reactors to burn plutonium stockpiles, use as a thermal breeder in conventional world dominant pressurized water reactors (PWRs); and due to its higher stability, in future applications of HTR fuel configuration. The results found in this review on identified advantages and challenges of thorium were mostly agreed with many authors cited. ii

3 TABEL OF CONTENT Page List of Tables..v List of Figures...vi Glossary of Symbols and abbreviations..vii Acknowledgements...ix CHAPTERS 1. Chapter 1 Introduction Background Scope and objectives of the study Methodology and organization of the dissertation Chapter 2 - Historical and technical characteristic of thorium- based nuclear reactors Early thorium-based nuclear reactors research and development The Radkowsky Thorium Reactor (RTR) Concept Incentives and justifications for thorium-based fuel cycle Chapter 3 - Characteristics of thorium fuel cycle Thorium fuel cycle and nuclear systems Thorium Thorium-based fuel cycles.12 iii

4 3.2 Thorium-based reactors and technologies Thorium fuel and nuclear systems Thorium fuel and breeding Reduced radioactive waste Thorium fuel cycles and plutonium burning Proliferation & radiochemistry aspects of thorium fuel cycle Proliferation and thorium fuel cycle Spent fuel processing, radiotoxicity and decay heat Chapter 4 - Viability of thorium-based nuclear reactors Advantages and disadvantages of thorium-based fuel cycles Thorium and Sustainability of Nuclear Energy Chapter 5 - Thorium fuel and emerging advanced reactor concepts Nuclear applications and potentials of thorium-based reactors Thorium fuel in current innovative and evolutionary nuclear systems Chapter 6 - Summary of Results Chapter 7 - Conclusion and Recommendation References Appendices...47 iv

5 LIST OF TABLES Pages Table 1 Table 2 Table 3 Table 4 Table 5 Table 6 Table 7 Some prototype thorium-based reactors earlier operated Some important characteristics of 3 basic fissile nuclei 10 World thorium resources 11 Production of Minor Actinides in Uranium and Thorium Fuel Cycles in g/t of heavy metal at 60 GWd/T...19 Critical mass for different plutonium compositions...23 Probability of an indicated yields Dose rate (mgy/h) from 1 kg of 233 U at a distance of 30 cm Appendix Table 8 Summary of experimental and power thorium based fuels reactors in the past.i Table 9 Decay heat emissions for different plutonium compositions... II Table 10 Properties of dominant plutonium and americium isotopes II Table 11 Isotopic composition of various grades of plutonium.ii Table 12 Plutonium produced in UO 2 and ThO 2 UO 2 cycles..ii Table 13 Peak temperatures at the Pu-high explosive interface.ii Table 14 Cost comparison using current prices III Table 15 Cost comparison using higher uranium prices...iii v

6 LIST OF FIGURES Pages Figure 1 World energy consumption and anticipated demands with projected contribution of likely energy sources....1 Figure 2 Thorium-based nuclear fuels designs (RTR concept)..8 Figure 3 Production of 233 U from 232 Th Figure 4 Neutron yield per neutron absorbed Figure 5 Number (n) of neutrons available for breeding in the uranium plutonium and thorium-uranium fuel cycles with thermal and fast neutron spectra. Breeding is impossible for negatives values of (n)...17 Figure 6 Available neutrons..18 Figure 7 Radiotoxicity of the actinides wastes.20 Figure 8 The overlapping life cycles of the different reactor generations 34 vi

7 GLOSSARY OF SYMBOLS AND ABBREVIATIONS? (eta) ratio of neutron yield per fission to neutrons absorbed? number of neutrons emitted by fission a the ratio of capture cross-section divided by the fission cross-section induced by neutrons as a function of energy s neutron cross-section ADS Accelerator driven systems CO 2 Carbon dioxide CRP Co-ordinated research projects GIF Generation IV International Forum HEU Highly enriched uranium HTGRs High temperature gas cooled reactors IAEA International Atomic Energy Agency IM Inert matrix INPRO Innovative Nuclear Reactors and Fuel Cycles Programme LMFBR Liquid metal cooled fast breeder reactor MICANET Michelangelo Network Competitiveness and Sustainability of Nuclear Energy in EU LWBR Light water breeder reactor LWR Light water reactor MA Minor actinides MOX Mixed oxide fuel MOX-T Mixed oxide with thorium fuel LWRs and HWRs (HEU as a fissile and thorium as a fertile) MSBR Molten salt breeder reactor MSR Molten salt reactor NEA Nuclear Energy Agency NPT Non-Proliferation to Nuclear Weapons Treaty PHWR Pressurized heavy water reactor PWR Pressurized water reactor RTR Radkowsky Thorium Reactor UK United Kingdom of Great Britain and Northern Ireland UN United Nations USA United States of America WNA World Nuclear Association vii

8 To my wife, Meseret for taking all the burden alone, and for my sons, Kidus and Redeate whom I missed them too much and which I always felt that I left them alone in the world when they need me at most. viii

9 ACKNOWLEDGEMENTS I would like to thank my project supervisor and course director, Dr. Paddy Regan, for his assistances and follow up. I acknowledge the course administrator, Mrs. Jayne Speed and other staff of physics department of University of Surrey for their assistance. I would also like to acknowledge in particular the International Atomic Energy Agency (IAEA) and ERPA for their support provided to me to attend the course. ix

10 CHAPTER 1 Introduction 1.1 Background Trends in the world's population and energy use during the past century show dramatic and relatively parallel increases in both [1]. These trends are expected to continue in the near future, led by growing demand in Asia. The demand for electricity is expected to increase more rapidly than the demand for other forms of energy throughout the world and nearly double by 2020 [1]. In its reference scenario up to the year 2050 presented in 1998, the World Energy Council also shows that if world primary energy demand doubles over the next 50 years, all energy sources will have to increase their contribution dramatically as shown in figure 1 below [2]. Figure 1. World energy consumption and anticipated demands with projected contribution of likely energy sources (WNA, [2]) This worldwide pressing energy demand, primarily from fast growing economies, nations like China, India, Brazil, South Africa and others; the rising concern of climate change and the need for clean sources of energy than traditional burning of fossil fuels that give rise to high-level of greenhouse gases like CO 2 ; the technical requirement of substantial amount of heat for fabrication of glass, cement and steel, or production of oil and hydrogen; desalination of water and concern of energy security of some states that are highly dependent on others lead to the need for production of significant amount of safe and clean energy which in return favours to nuclear options. 1

11 Other renewable alternative sources of energy like solar and wind can assist to some extent but current advancement indicated that they are in short of supplying the required growing demand of energy either economically or/and in substantial amount [3]. The nuclear option can deliver considerable amount of electricity and heat production to the world. The heat generated from nuclear power plants is of sufficiently high temperature (>700 0 C) and could be used for coal gasification or production of hydrogen, in which the hydrogen-based energy system can then progressively replace the carbon-based fuel mainly in the transport sector and minimize higher emission of CO 2 [4]. The relatively lower temperature can also be used for applications like district heating, desalination of potable water, etc [4]. Even though nuclear energy provides many rewards to the wellbeing of people; and nuclear power plants have a relative good record of safety; many people remain unenthusiastic towards it and have worries for its expansion. This could be mainly accounted for lack sufficient information and the non-transparent activities of some nuclear research and development programs. The concern of proliferation to nuclear weapons and the fate of radiotoxic waste generated are still key issues on the nuclear agenda. The other crucial concern of many people is also what will happen if it falls in the hands of extremists. This in return also leads to highest security system which will further pressurize the nuclear industry; even though safety and security measures overlap to some good extent. Recent news from the IAEA also indicated that more developing countries new to nuclear technology, such as Saudi Arabia, Egypt, Jordan, Vietnam, Nigeria, are lining up the nuclear option to unravel their energy demands and have requested the IAEA for technical assistances. Other motives of nations to diversify their energy sources to nuclear options could be: to deal with natural disasters (for instance effect of draught in hydroelectric dams); decline of national fossil fuel reserves; and the strategic utilization of national resources. Though numerous, the reasons may be it seems now there is growing interest of nations (mainly from developing countries) to nuclear power, which in return may raise concerns of the proliferation risk associated with nuclear weapons and safety issues. These concerns in some instances become complicated and can developed to embittered international affairs, such as the current relationships between the USA and the Iranian and North Korean regimes. There is also an equally challenging tenet of the early atom-for-peace spirit of the world that preserves the legal right and responsibilities of all nations to make use of peaceful applications of nuclear technology for their development that respect the NPT principles, which provide an assurance about exclusively peaceful use of nuclear material and facilities. 2

12 It is worthwhile to mention here that it is not only electricity and heat nuclear technology offers to the world but many diverse social and economic rewards that can improve the welfare of people, from killing cancerous tumors to pest control, and understanding the mysterious nature of elementary particles. Therefore, it is sensible that nations show great interest towards it. Yet, it should be, as with all responsibilities that the technology entails and a full commitment of states for its peaceful applications as well as developing all other prudence capabilities that the technology requires. Therefore, to reconcile this impediment, i.e. to make use of the nuclear technology for the advancement and benefit of people and avoid the threat of safety and proliferation issues requires a careful and responsible approach. The difficulties in managing radioactive waste generated from reactors should also be properly addressed. Thus, the ultimately solution may fall on a robust design of nuclear reactors and fuel cycles which are inherently safe and proliferation resistant with genuine responsibility of all nations that need the technology, maintained by effective implementation of international safeguards regime. In this regard, in addition to long term considerations of world energy sustainability; the thorium-based fuel system offer a real promise of unproved performance over all-uranium LWR reactors, which seem to be approaching their economic limits on burnup and have foiled attempts to press on further into the long refueling cycle, ultrahigh-burnup operating regime [1]. 1.2 Scope and objectives of the study The aim of this review is therefore to cover the feasibility and place of thorium-based fission reactors and its fuel cycle to answer the challenges facing the existing growing global energy demand, and discuss the past and present efforts to harness its benefit including future nuclear applications in sustainable way. The main objectives of this review are: To examine from literature survey the viability (advantages and disadvantages) of thorium-based fission nuclear reactors; features of thorium fuel cycle and its potential in future applications. To analyze historical and technical characteristics and the radiochemistry aspects of thorium-based reactors and its fuel cycle compared with nuclear proliferation. 3

13 1.3 Methodology and organization of the dissertation The results of much literature from past and current research and development activities on thorium-based fuel cycles has been critically reviewed, analysed and presented in a concise way. Due to its nature there may be limitations in obtaining more detailed information from companies and research establishments. Furthermore, useful information found from the IAEA Co-ordinated Research Projects (CRP) on thorium-based fuel systems has been used. In this review, chapter 2 describes mainly early thorium-based nuclear reactors research and development activities, early historical and technical concepts and challenges, and incentives and justifications to use thorium-based fuel cycles. Chapter 3 present abundance and relevant features of the element thorium as reactor fuel; an evaluation of the status of thorium-based fuel cycles and nuclear systems; breeding and fuel reprocessing; and an evaluation of the toxicity of thorium fuel cycle waste compared to other fuel cycles. New incentives to use thorium fuel cycles due to large stockpiles of plutonium produced in nuclear reactor; and the radiochemistry & proliferation aspects of thorium fuel cycles are discussed. Chapter 4 examines the viability (advantages & disadvantages) of thorium-based nuclear reactors; and contribution of thorium-based fuel cycles to sustainability of nuclear energy. Chapter 5 presents the role of thorium fuel in emerging innovative and advanced fuel cycles and reactor concepts and focus on evaluation of nuclear applications and potentials of thorium-based reactors. Finally, in chapters 6 and 7 the summary of results and conclusion of the review are presented. 4

14 CHAPTER 2 Historical and technical characteristic of thorium-based nuclear reactors The interest in thorium fuel cycles has been revived in recent years due to the recognition of "new" potential benefits, principally the potential to reduce plutonium and minor actinide production [19]. In addition, better material properties and fuel behavior increase the attractiveness of thorium-based fuel cycles. In this chapter early efforts and studies of thorium-based fuel reactors and its features; basic technical information and concepts that are used to utilize thorium fuel cycles and the existing incentives and justifications to realize a thorium-based fuel cycle are discussed. 2.1 Early thorium-based nuclear reactors research and development Many reference materials indicated, study of thorium reactors started in the early days of nuclear technology developments [1, 3, 4, 6, 8]. For example, in the USA as a follow-up of the Manhattan Project ( ), the leading US laboratories singled out thorium as a possible complement for uranium and the possibility of using 233 U as a nuclear weapon [5, 6]. Lung and Gremm [5] and Unak [6] also reported that in 1958, about 55 kg of 233 U were produced and available in the US. In their paper on perspectives of thorium fuel cycle, Lung and Gremm [5] prompt a notable early consideration of thorium fuel that It is interesting to recall the predictions of the International Fuel Cycle Evaluation Conference (INFCE) of 1978 (published in 1980 by IAEA), where thorium was given almost equal importance as uranium. However, as most agrees, due to mainly economical and strategic factors priority was given to uranium than thorium [6]. Nevertheless, technical studies on the use of thorium nuclear fuel cycle were slowly progressed during the last 50 years; but serious technical progresses were realized on the uranium and plutonium fuel elements [6]. During this gradual thorium-based fuel cycle emergent period, thorium extraction processes were developed in some countries and about 1500 kg of 233 U have been separated in the USA from only 900 tones of thorium [5]. It has been also reported [5, 6] that USA and France have each separated about 2000 tones of thorium during the period of , and part of which is still available. In addition, it was noted that many reactor prototypes were built and operated and thorium extraction plants were set up in many countries [6, 7, 10]. 5

15 Table - 1 below summarizes some thorium-based prototype reactors that were operated earlier. (It was observed that some of the operation periods and output powers provided vary in the referred citations [6, 7, 10]; more detailed information is also provided in Table-8 Appendix- I). Table -1: Some prototype thorium-based reactors earlier operated [6, 7]. Location or Name Operation time Country Power (MW) Shippingport USA 100 Indian-Point USA 285 Elk-River USA 24 ORNL USA 7.5 Peach Bottom USA 40 Dragon UK 20 AVR Germany 2.4 SUSPOP Holland 1 Fort St-Varen USA 300 THRT Germany 300 In these earlier research and development activities, one of the main challenges encountered by many in the design of a thorium fuel based systems is the necessity to supplement natural thorium with a pregenerated fissile component or a source of neutrons [8]. To overcome this challenge it was proposed to use particle accelerators to generate neutrons, but this process is costly, and the other practical means is to combine the thorium with conventional nuclear fuels (made up of either plutonium or enriched uranium or both) the fissioning of which provides the neutrons to start breeding and further reactions [9]. In its effort in summarizing and documenting past experiences on thorium fuel cycle, the IAEA [10] reported that basic research and development activities has been conducted in Germany, India, Japan, the Russian Federation, the United Kingdom, and the USA. These studies included the determination of material data, fabrication tests on a laboratory scale, irradiation of Th-based fuel in material test reactors with post-irradiation examinations, and investigation to the use of Th-based fuel for LWRs (including WWERs), LMFBRs, and HTRs/HTGRs. One of these important activities to realize the thorium-based fuel cycle in proliferation- resistant manner, to adapt in the world dominant PWR is the Radkowsky Thorium Reactor (RTR) concept that will be discussed next with special considerations. 6

16 2.2 The Radkowsky Thorium Reactor (RTR) Concept Unak [6] in describing the potential use of thorium, mentioned that, it is promising progress that a new thorium-based nuclear fuel proposed by Radkowsky as being based on a seed-blanket fuel design looks an ideal approach to replace the uranium based nuclear fuels. He [6] also remarked Starting from this new fuel approach Radkowsky Thorium Power Corporation has been established in the USA in Galperin, et al [8] remarked that the RTR concept proposed by Professor A. Radkowsky offers a solution to the thorium utilization problem. The basic idea is to use the heterogeneous, seed/blanket (SBU), fuel assembly [8]. This separation allows separate fuel management schemes for the thorium part of the fuel (a subcritical blanket ) and the driving part of the core (a supercritical seed ). The design objective of the blanket is efficient generation and in-situ fissioning of the 233 U isotope, while the design objective of the seed is to supply neutrons to the blanket in the most economic investment of natural uranium [8]. In unfolding the achievement of the RTR fuel cycle, Galperin, et al [8] further discussed that the SBU geometry provides the necessary flexibility to satisfy a major design constraint; i.e. full compatibility with existing pressurized water reactor (PWR) power plants. In addition to that the heterogeneity of SBU design allows the necessary (and separate) optimization of seed and blanket lattices [8]. Kazimi [9] also further described the efforts by another program on the objective to look at practical ways to simplify the RTR fuel design of the separated seed and blanket units remarked that Investigators at Brookhaven National Laboratory together with the Center for Advanced Nuclear Energy Systems (CANES) at Massachusetts Institute of Technology (MIT); although the terminology of seeds and blankets has been kept (with the new name of this arrangement as the whole-assembly seed-and-blanket core), the metaphor is less applicable in this case, which calls for these assemblies to be arranged, more or less, in a checkerboard array within the core of a reactor. This new whole-assembly seed-and-blanket core arrangement with RTR fuel matrix is shown in figure -2 below. 7

17 Figure-2 Thorium-based nuclear fuels can be designed in different ways. One general scheme, first conceived by Radkowsky, is to have each nuclear fuel assembly (squares) composed of uranium-rich "seed" rods surrounded by thorium-rich "blanket" rods (top). The uranium, which includes up to 20 percent of the fissionable isotope 235 U, produces enough neutrons to transform the "fertile" thorium around it into another fissionable isotope of uranium, 233 U. This mixing of fuel types within an assembly complicates the refueling of a nuclear reactor, because the seed rods need to be replaced much more frequently than the blanket rods. An alternative approach, called the whole -assembly seed-and-blanket core (bottom), utilizes fuel assemblies that each contain only uranium-rich seed rods or thorium-rich blanket rods. These assemblies can be more easily shuffled or replaced at prescribed intervals. (Barbara Aulicino) [9] 8

18 Galperin, et al [8] also stated that in the RTR fuel cycle, because of fuel economy reasons, the Th-blanket which is about 60% of the fuel assembly, in-core residence time is quite long (about 10 years), while the uranium part of the SBU which is about 40% of the fuel cycle, is replaced on annual (or 18 month) basis, similar to standard PWR fuel management practice. Here, the long blanket residence time is required to achieve very large accumulated burnup of the thorium part of the fuel, about 100 GWd/t (10 GWd/t on average for each annual cycle). A detailed evaluation of the RTR fuel concept has also been done by Paul R. Kasten [20]. He provides thorough comparisons of his findings with the claims made by the RTR concept and concluded that Overall, it appears that the RTR has desirable non-proliferation features and provides a novel way to achieve high fertile material burnup in PWRs. Further, for the reference evaluation conditions, the fuel cycle of cost of the RTR was slightly less than for a PWR, but the difference was small compared with the uncertainty in cost parameters. By displacing the thorium with natural uranium in the RTR concept, the fuel cycle cost might be further reduced while retaining desirable non-proliferation features. However, the practicality of the concept requires additional R&D to achieve the blanket and seed fuel performance required, and demonstration is required of the practicality of the fuel shuffling/reloading occurring during refuelling. Thus, considering the advantages of the RTR fuel cycle and the improved whole-assembly seed-and-blanket concept with its potential benefits, among others, longer burnup time, less waste generation, and its resistance to weapons proliferation; RTR fuel can be one of the thorium-based fuel cycles that can be deployed in near future. This in turn can make use of thorium reserves as fuel to sustain nuclear energy further. As a result, the RTR fuel cycle warrants sufficient investment and coordinated effort to conclude the existing drawbacks at some acceptable levels. 2.3 Incentives and justifications for thorium-based fuel cycle It was found earlier [6, 8, 10] that 233 U presents a superior fissile nuclide producing more neutrons per thermal neutron absorbed than all other fissile isotopes. Table 2 below shows the comparison of fission properties of 233 U, 235 U, and 239 Pu as being most important fissile nuclei in reactors. It is shown that 233 U is a nucleus having the highest fission factor and good fission cross-section for thermal neutrons [6]. These basic fissile characters of 233 U mean that it can be used as a fuel in nuclear reactors. So, the natural 232 Th as a breeder of 233 U has became an important nuclear material having very large potential applications [6]. 9

19 Table 2 Some important characteristics of 3 basic fissile nuclei [6]. Fissile nucleus Fission factor (µ) Neutron yield (?) Fission cross-section (s f, barn) 233 U U Pu Galperin, et al., [8] and Unak [6] assert that, this features of 233 U, and the fact that thorium is much more abundant as a natural ore than uranium motivated numerous attempts to design and implement a nuclear reactor based on thorium fuel. The most notable examples are the Light Water Breeder Reactor (LWBR) and early High Temperature Gas Cooled Reactor (HTGR) [8]. It was also remarked [4] that the feasibility of thorium utilization in high temperature gas cooled reactors (HTGRs), light water reactors (LWRs), pressurized heavy water reactors (PHWRs), liquid metal cooled fast breeder reactors (LMFBR) and molten salt breeder reactors (MSBR) were also demonstrated. Herring et al., [1] identified the current challenges of nuclear energy and summarized it; in a more concise way that to make use of nuclear option it is necessary to develop advanced reactors and fuel cycles: that operate at less cost per MW day; which allows longer refueling cycles and higher sustainable plant capacity factors; that is resistant to nuclear weapon-material proliferation; which results in a more stable and insoluble waste form; and, that generates less high-level waste. In light of these challenges and current rising energy demand, the following part of this review discusses the role of thorium-based fuel in meeting the next generation requirements. 10

20 CHAPTER 3 Thorium fuel cycle and nuclear systems The fuel to nuclear power plants so far mainly comes from uranium either natural which is a mix of large proportion of the fertile 238 U (99.3%) and the fissile component 235 U (0.7%) or enriched uranium with different enrichment levels of 235 U depending on the design of the reactor used. Another natural source of fuel to reactors is thorium. This chapter mainly discusses types and nature of thorium-based fuel cycles including a brief description of the element thorium, and thorium s role in proliferation-resistance and waste reduction. 3.1 Thorium as nuclear fuel Thorium, like uranium, can be used as a nuclear fuel. Although not fissile itself, 232 Th will absorb neutrons to produce 233 U, which is fissile. Hence like 238 U it is fertile Thorium Thorium occurs in several minerals, the most common being the rare earth-thorium-phosphate mineral, monazite, which contains up to about 12% thorium oxide, but average 6-7% [7]. Soil commonly contains an average of around 6 ppm (parts per million) of thorium [7]; others [4] reported that thorium content of soil as much as 10 ppm. However, it was agreed that there are substantial deposits in several countries (see table 3 below) and thorium is estimated to be three to four times more abundant than uranium [4]. Table 3 World thorium resources (economically extractable): Country Reserves (tonnes) Australia India Norway USA Canada South Africa Brazil Other countries World total Source: US Geological Survey, Mineral Commodity Summaries, January [7] 11

21 232 Th decays very slowly (its half-life is about three times the age of the earth 1.4x10 10 years) but other thorium isotopes occur in its and in uranium's decay chains. Most of these are short-lived and hence much more radioactive than 232 Th, nonetheless considering their mass they are negligible [7]. Thorium oxide, the form of thorium used for nuclear power also called thoria, has one of the highest melting points of all oxides (3300 C) [7]. It is a highly stable compound than uranium dioxide, which are currently used in reactor fuels. Therefore, there is less concern that fuel pellets could react chemically with the metal fuel cladding around them or with the coolant water if there should be a breach in the protective cladding [7]. Kazimi [9] in describing the advantages of thorium fuel as compared with the uranium dioxide fuel cycle also reported that the thermal conductivity of thorium dioxide (ThO 2 ) is 10 to 15 percent higher than that of uranium dioxide (UO 2 ), making it easier for heat to flow out of fuel rods in the reactor. Furthermore, he commented that, the melting point of thorium dioxide is about C higher than that of uranium dioxide, and this difference provides an added margin of safety in the event of temporary power surge or loss of coolant. Therefore, considering its important features as fuel to reactors and its relative ample abundance, thorium can be a good supplement, even better in some instances, to the current uranium reserve to sustain nuclear energy further for some centuries ahead Thorium-based fuel cycles L.C. Walters, et al. [15], in their paper on nuclear fuel considerations for the 21 st century, forecast among others, the importance of deployment of thorium fuel cycle in the coming next 50 years and noted that several events, such as fear of the consequences of the greenhouse effect and the resulting carbon tax, may lead to nuclear power becoming economically attractive. They remarked, This, in turn, would increase the rate at which uranium reserves diminish due to the increased rate of nuclear power deployment. However, breakthroughs in the extraction of uranium from the sea or deployment of fast breeder reactors would greatly extend the uranium reserves. Development and implementation of a thorium fuel cycle would also greatly extend the nuclear power enterprise. However, beyond this concern of resource sustainability, there are many superior advantages of using thorium as nuclear fuel such as better neutronic property of 233 U. Given a start with some other fissile material ( 235 U or 239 Pu), a breeding cycle similar to but more efficient than that with 238 U and plutonium (in slow-neutron reactors) can be set up [8]. 12

22 This is mainly because of the thermal absorption cross-section of 232 Th (7.4 barns) is three times higher than that of 238 U (2.7 barns), though the resonance integral of 232 Th is one-third of that of 238 U [4]. The 232 Th absorbs a neutron to become 233 Th which normally decays to 233 Pa and then 233 U. The irradiated fuel can then be unloaded from the reactor, the 233 U separated from the thorium, and fed back into another reactor as part of a closed fuel cycle [7]. Production of 233 U from 232 Th is shown schematically in figure 3 below. Figure 3 Production of 233 U from 232 Th [1] The advantage of thorium in Light Water Breeder Reactor (LWBR) concept has also been emphasized by some authors [1, 7, 11]. It was successfully demonstrated in the USA in the 1970s that major potential application for conventional PWRs involves fuel assemblies arranged so that a blanket of mainly thorium fuel rods surrounds a more-enriched seed element containing 235 U which supplies neutrons to the subcritical blanket; the 233 U produced in the blanket is also burned in the reactor. Another option to this heterogeneous fuel setup is the use of whole homogeneous assembles arranged so that a set of them makes up a seed and blanket arrangement [11]. If the seed fuel is metal uranium alloy instead of oxide, there is better heat conduction to cope with its higher temperatures. Seed fuel remains three years in the reactor, blanket fuel for up to 14 years [11]. Galperin, et al [8] in the RTR fuel concept also explained in detail the fuel burnup time and neutron deployment for efficient utilisation of thorium in a once-through cycle which encounters a neutron economy problem. In order to recover this investment in terms of fuel utilisation gains by taking advantage of superior 233 U properties, the thorium-based fuel 13

23 must be burned further, to a burnup of at least GWd/t, corresponding to 8-9 full-power years [8]. Thus, they [8] remarked that the main challenge of efficient utilisation of thorium in LWRs is reduced to a problem of achieving very large accumulated burnup of the thorium in a once-through fuel cycle. Another interesting thorium fuel reported by M. Osaka, et al. [16] is the americiumthorium oxide phase of the cermet fuel. According to the author [16], this ThO 2 -based Am-containing cermet fuel is composed of micro- or macro- spheres of (Am, Th)O 2-x dispersed in a molybdenum (Mo) inert matrix which is mainly due to its stability with the fluorite structure. The structure represents nuclear fuel mainly because of its structural advantages such as isotropy, high metal density, high melting temperature, and thermal conductivity. They [16] reported that ThO 2 based fuel indicates some neutronic advantages over UO 2 -based fuel, i.e. yielding a new fissile material of 233 U and suppressing the newly generating MAs by burnup. In summarizing thorium-based fuel types, in which a number of experimental and prototype power reactors were successfully operated during the mid 1950s to the mid 1970s the IAEA document [4] identified as: (Th, U)O 2 and (Th, U)C 2 fuels in high temperature gas cooled reactors (HTGR), (Th, U)O 2 fuel in light water reactors (LWR) and the flibe type salt (Li 7 F/BeF 2 /ThF 4 /UF 4 ) fuel in molten salt breeder reactor (MSBR). It was also reported here [4] that 232 Th and 233 U are the best fertile and fissile materials respectively for thermal neutron reactors. Besides, thermal breeding has been demonstrated for (Th, U)O 2 fuel in the Shippingport light water breeder reactor (LWBR) in the USA. ThO 2 has also been successfully used as blanket material in liquid metal cooled fast breeder reactor (LMFBR) and for neutron flux flattening of the initial core of pressurized heavy water reactor (PHWR) during startup. In examining thorium fuel types, there are open and closed thorium fuel cycles that can utilize 233 U converted from the fertile 232 Th. Open fuel cycles, such as the Radkowsky concept, are based on irradiation of 232 Th and in situ fission of 233 U, without involving chemical separation of 233 U. This cycle avoids the engineering processes and other complications associated with processing and fabrication of highly radiotoxic 233 U-based fuels [4]. The closed fuel cycle based on chemical reprocessing of irradiated thorium or thoriumbased fuels for recovery of 233 U and refabrication and recycling of 233 U bearing fuels. In this case, LWRs like WWER-1000 using mixed thorium-plutonium fuel can be considered. For recycling the 233 U an important factor is the 232 U content in 233 U. It was reported, for a 14

24 standard burnup of 40 MWd/kg HM for a WWER-1000 fuel, the 232 U content would be in the range of 3000 ppm [4]. The two recycling options are [4]; the use of ( 232 Th 233 U)O 2 fuel; and the use of (Depleted U 233 U)O 2 or (Reprocessed U from WWER 233 U)O 2. Generally, it can be observed that thorium fuel demonstrated many superior advantages as nuclear fuel such as better neutronic property of 233 U. Beside, thorium-based fuels can be run with all available current reactor types; experimental and prototype power reactors were also successfully operated. In addition, thorium-based fuel cycles found applications in advanced reactor concepts, such as HTGR and MSBR. 3.2 Thorium-based reactors and technologies Thorium fuel and nuclear systems Nuclear reactors produce energy by burning nuclear fuel. The nuclear fuel presents a complex engineering product and is a subject of extensive research and development studies. There are two main types of nuclear fuel cycle using thorium. In the first type Accelerator Driven Systems (ADS), a particle accelerator coupled to the reactor generates high-energy neutrons through the spallation reaction of high-energy protons striking heavy target nuclei (lead, lead-bismuth or other material) [11]. These neutrons can be directed to a subcritical reactor containing thorium, where the neutrons breed 233 U and promote fission further. Therefore, it is possible to sustain a fission reaction which can readily be turned off, and used either for power generation or destruction of actinides resulting from the U/Pu fuel cycle [11]. The other main type (the mixed fuel Thorium cycle), is the initial kick can be provided by using a small amount of fissionable Uranium (or plutonium). In this case, once given a starting fissile material ( 235 U or 239 Pu), the 232 Th absorbs a neutron and produce 233 U, another fissile material that can efficiently burn in a reactor. The mixed fuel thorium cycle has some distinct advantages over the uranium cycle, including that the resultant waste products are much smaller in volume than conventional uranium fuel cycle [17, 14]; for instance, in the case of MOX-T (HEU-Th) fuel, [19] quantities of 239 Pu and other long-lived actinides generated are at least 100 times less than in conventional fuel. Lung and Gremm [5] also after describing the advantage of thorium fuel from neutronic standpoint assert that the eta ratio (?) of neutron yield per fission to neutrons absorbed is higher to that of 235 U or of 239 Pu as shown in figure 4; and confirmed that 233 U will be a good fuel in any type of reactor. 15

25 Figure 4 Neutron yield per neutron absorbed (source: OECD-NEA, 1989). [5] A different reactor concept, of high power density (DT) fusion-fission (hybrid) reactor which is loaded with thorium fuel, was also explained by Ubeyli and Acir [3]. After remarking the easy abundance of fusion fuel in nature compared to the scarce fission fuels and predicting the realization of fusion reactor as, may not be available commercially in the coming 50 years, they [3] described, early introduction of fusion nuclear systems is possible by fusion-fission hybrid reactor and fuel concept as a multi-functional reactor type that is a combination of the fusion and fission processes. The main purpose of this type of hybrid reactor is to convert fertile materials ( 238 U or 232 Th) surrounding the fusion plasma into fissile materials ( 239 Pu or 233 U) by transmutation through the capture of high yield fusion neutrons and to amplify energy as well as burning and/or transmutation of nuclear waste effectively [3]. One other important aspect that warrant due consideration is the availability of experience to realizing a technology. In this regard, in manufacturing of fresh thorium fuel, Vapirev et al., [19] assert, There is considerable experience with the thorium-containing fuels 233 UO 2 -ThO 2 and 235 UO 2 -ThO 2. Thorium-containing fuels have been produced for the Indian Point I and Elk River reactors, the Borax IV BWR and also for the LWBR at shippingport, which has a blanket of 233 UO 2 + ThO 2 and ThO 2 in the shape of pellets. There is experience also with U-Th fuel at the prototype plant in Saluggia, Italy and also in Germany both with water reactors and HTGRs. Most of the experience up to now is with U-Th fuel for HTGRs. 16

26 It was found that different types of thorium-based fuel concepts and nuclear systems existed; of these some are tested and others are in developing stage. Besides, it is observed that there are experiences in fuel manufacturing of fresh thorium fuels in some countries which is an important step in utilizing thorium fuel cycles. Some of advanced emergent thorium-based fuel concepts such as molten salt reactors will be further discussed later Thorium fuel and breeding One unique significant value of thorium-uranium fuel cycle ( 232 Th 233 U) is its breeding by capture of whether thermal or fast neutrons by 232 Th, while the uranium plutonium ( 238 U 239 Pu) fuel cycle only requires fast neutrons to be sustainable, figure 5 below shows [13] the available number of neutron for breeding of the two fuel cycles in slow and fast spectrum. Figure - 5 Number (n) of neutrons available for breeding in the uranium plutonium and thorium-uranium fuel cycles with thermal and fast neutron spectra. Breeding is impossible for negatives values of (n). [13] David, et al., [13] remarked that the availability of large amounts of plutonium led to the development of the so-called U-Pu fuel cycle; besides they assert that the need for breeding reactors was conceived at earlier times to sustain nuclear energy in the years ahead. Le Brun [17] in his evaluation on thorium fuel to deploy in molten salt reactor concept also defined and explained the characteristics of breeding in thorium and uranium fuel cycles in range of neutron spectrum as, a reactor will be a breeder if it is able to produce a number of fissile nuclei equal or greater than the number of fissile nuclei that disappear by capture or fission. If? is the number of neutrons emitted by fission and a the ratio of capture 17

27 cross-section divided by the fission cross-section induced by neutrons as a function of energy, the available neutrons left by the breeding is given by N b =? 2(1 + a). This quantity is plotted in fig. 6 below for the two fertile elements. It appears that, as the available neutron number is slightly larger than 0, breeding is possible for the whole neutron energy spectrum for thorium whereas it is only possible for neutron energy larger than a few 10 kev for plutonium. This explains why, if plutonium is produced and partly burnt in the light water reactors, it is impossible to reach desirable breeding ratio with a thermal neutron spectrum in the U/Pu cycle. He [17] further concluded that the main advantage of the thermal spectra is that the required fissile material for starting the chain reaction is smaller (factor up to six) than for the fast neutron reactor, so the deployment is easier. Energy (ev) Figure - 6 Available neutrons [17] Where,? neutrons are produced 1 neutron induces a new fission a extra neutrons are captured by fissile (1 + a) neutrons are captured by fertile for fissile regeneration This required a small initial inventory of thorium fuel cycle in breeder reactor compared to uranium cycle again confirmed by David et al., [13] as, the production (by 232 Th irradiation in a classical reactor) of the initial 233 U load for a thorium-uranium breeder reactor would be four times shorter than the production (by 238 U irradiation in the same type of reactor) of the initial 239 Pu load for a uranium-plutonium breeder reactor. In general, breeding exist in all neutron spectrum of thorium fuel and the initial 233 U load to thermal thorium-uranium breeder reactor is smaller than a fast U/Pu reactor. In addition, due to higher probability of fission in thermal neutrons than fast spectrum, thorium-based fuels show as a best breeder of fissile material ( 233 U) in LWRs thermal reactors. 18

28 3.3 Reduced radioactive waste One important issue and public concern in utilizing nuclear energy is the issue of radiotoxic waste generated. Therefore, to sustain nuclear power plants in the coming years, there is a need to find a means to manage spent fuels and minor actinides (MAs) generated from nuclear systems. To realize this, due attention, should be given at the initial phase of designing and opting fuel cycles. One current driving force to make use of thorium-based fuel cycles is; therefore, its potential in minimizing radiotoxic waste and long-lived MAs generated, and its potential in transmuting MAs to acceptable levels. In this regard, Sundararajan et al., [14] assert, the greatest advantage of thorium fuel cycle lies in the fact that it produces significantly less quantities of long lived minor actinides than the uranium fuel cycle as it can be seen from Table - 4 below. They claim that the radiotoxicity of the transuranics in the spent fuel of conventional LWRs is about 100 times higher than that in the spent fuel from 232 Th- 233 U fuel cycle; they further commented, the two nuclides which are of environmental concern and which are significantly present in the waste streams are 231 Pa and 237 Np; in which the behaviour of these radionuclides in the process streams and their transport characteristics in the environment need further investigations. And they concluded, once through mode of fuel cycle is adopted, thorium oxide materials are likely to be more enduring than would be the case with uranium. Besides, they [14] claimed that the high degree of chemical stability and very poor solubility of thoria make spent thoria based fuels like (Th-Pu)O 2 ideal as waste forms for direct geological disposal. Table 4 Production of Minor Actinides in Uranium and Thorium Fuel Cycles in g/t of heavy metal at 60 GWd/T [14]. Different Fuel Compositions Minor Actinides 235 U U 235 U Th 233 U U 233 U Th 237 Np 9.0E E E E+01 Am 4.7E E E E-03 Cm 2.2E E E E-03 Le Brun [17] in evaluating thorium fuel to deploy in molten salt reactor concept also observed that another desirable feature of the thorium cycle is the lower production of actinides which are the main contributors to the radiotoxicity of spent fuel. He [17] further 19

29 argued that the radiotoxicity which tries to assess the risk due to the spent fuel of the various fuel cycles plotted as a function of time is given in figure - 7 below, which shows clearly the advantage of the thorium cycle. Time (y) Figure - 7 Radiotoxicity of the actinides wastes [17] To give an idea of the level of the curves presented in figure - 7, he explained as, the dose associated with the natural uranium which would be needed to produce one GWe*Year is 5x10 5 Sv. The radiotoxicity of the spent fuel coming from a PWR and stored without reprocessing is given by the upper blue curve; it reaches the original and natural radiotoxicity of uranium only after 107 years. With the U/Pu cycle in fast neutron reactors, the produced radiotoxicity is reduced by an order of magnitude (middle red curve). Another order of magnitude is still obtained with the thermal thorium cycle operating in molten salt reactors (lower green curve). The radiotoxicity produced by the fission products is the same whatever the cycle and is given by the dashed curve. This comparison of radiotoxicity of the transuranics in the spent fuels of Th-based and PWR fuel cycles made by Le Brun [17] is also in agreement with the above claim made by Sundararajan et al., [14]. Another most important property that affects almost all aspects of fuel behaviors of oxide fuels is the oxygen potential [16]. In handling waste, the fuel reactivity with oxygen should be investigated thoroughly. In this respect, Herring et al., [1] has compared the relative proportions of thoria and urania that take part in a chemical reaction assert that thorium dioxide is the highest oxide of thorium and does not depart significantly from its stoichiometric composition of ThO 2 when exposed to air or water at temperatures up to K. They conclude that ThO 2 -UO 2 mixtures appear to be susceptible to corrosive attack in air or oxygenated water, but significantly less susceptible than UO 2. 20

30 Quoting, the investigation made by Markowitz and Clayton (1970) on the corrosion behaviour of a group of nuclear fuel oxides, Herring et al., [1] consolidated the above findings as, The ThO 2 samples displayed excellent corrosion resistance. The weight gains of the urania-thoria material exposed to water oxygenated to about 100 ppm were much larger than for any of the other materials tested, in any of the media, even those which faded and fell apart; yet all the specimens retained their mechanical integrity. The urania-thoria fuel remained mechanically intact in spite of the growth of an oxidized surface phase thought to be of the M 4 O 9 type They concluded [1]; therefore, a mixture of ThO 2 and UO 2 is much more resistant to long-term corrosion in air or oxygenated water than UO 2. Thus ThO 2 -UO 2 is a superior waste form if the spent fuel is slated for direct disposal rather than reprocessing. Vapirev et al., [19] also remarked the environmental impact due to radioactive waste from thorium mining is small as compared to that of uranium as long as if it is not accompanied by uranium. They justified this findings by stating, The longest-living isotope of 232 Th daughter products are 228 Ra (6.7 years) and 228 Th (1.9 years), i.e. if 232 Th is extracted, the tailings will decrease their activity 20 times in 30 years. 3.4 Thorium fuel cycles and plutonium burning Conversion of the existing stockpiles of weapon-grade high enriched uranium (HEU) is a problem that has emerged in recent years. According to estimations, the 50, 000 warheads in the United States and Russian arsenals contain some 1000 t of HEU and 220 t of plutonium. This data and concern was taken from Vapirev et al., [19] paper on the possibility of using weapon-grade HEU and plutonium in thorium oxide fuel (MOX-T) for LWRs and HWRs. They [19] remarked, to resolve this problem that the most straight forward solution for conversion and utilization of HEU is to blend it with natural uranium; and the argument in favour of blending is that the process is irreversible and such technology promotes observation of the NPT. However, an argument against this solution mentioned that a lot of energy has been used for the enrichment; therefore, mixing HEU with natural uranium has to be considered as the last option; another problem of the contemporary nuclear fuel cycle is also the inventible generation of 239 Pu and other higher long-living actinides. 21

31 Therefore, the proposed new fuel cycle is to use 235 U (HEU) as a fissile isotope and 232 Th as a fertile isotope in a mixed oxide with thorium (MOX-T) fuel for light and heavy water reactors (LWRs and HWRs). Plutonium can also be used in the proposed fuel along with 235 U. They [19] recalled some features of HEU fuel topped with thorium have been discussed in a Report on Advanced Fuel Cycle and Reactor Concepts (INFCE, 1980). As claimed by the author [19], the main advantages of MOX-T fuel in burning HEU (Pu) can be summarized as: i. 239 Pu and other long-lived actinides are generated in quantities which are at least 100 times less than in conventional fuel; ii. Neutron emission is lower by a factor of more than 100; iii. 233 U is generated and burnt (the conversion factor for LWRs is and for HWRs about 0.88); iv. Thorium is utilized and the total available amount of nuclear fuel is increased. Lung and Gremm [5] in describing the contribution of thorium fuel cycle in burning of excess plutonium also assert that The bad image of plutonium is not based on scientific arguments. Therefore, it should not be regarded as a waste product, but as a valuable fissile material. A good use of plutonium would be to use it as a tool to achieve sound reactivity in a thorium facility. 3.5 Proliferation & radiochemistry aspects of thorium fuel cycle Proliferation and thorium fuel cycle After discussing the handling and managing of spent fuel in reactors as an important process, Galperin et al., [8] described the proliferation issue associated with spent fuels as, for nuclear power to be accepted as a main source of energy in the next century, it must be based on a fuel cycle which is highly proliferation-resistant. They concluded that The non-proliferation nature of the nuclear power fuel cycle material flow should be supported not only by a combination of administrative safeguard measures, but mainly by avoiding production of any material of such quantity and quality as to be potential weapons use in the fuel cycle itself. Besides, they remarked, an additional factor is the measure of complexity required to separate the fissile component from the normal material flow of the fuel cycle [8]. 22

32 The author [8] recalling an earlier review on proliferation and its conclusion stated that, The extensive Nonproliferation Alternative Systems Assessment Program (NASAP) studies concluded in 1980 that none of the existing or proposed fuel cycle schemes were immune to the possibility of proliferation. Due to the fact that the main proliferation potential is associated with plutonium (Pu), created by transmutation of 238 U, thorium presents a natural alternative fertile material. Thus, they emphasised thorium fuel can be a promising proliferation-resistant than 238 U/Pu cycle as a future energy source. In case of deliberate diversion of nuclear materials from civilian nuclear power fuel cycles by nations, they [8] further argued that International treaties and safeguards may not be effective because they can be abrogated or circumvented; therefore, international safeguards are necessary but not sufficient. An important barrier to proliferation is therefore should be based on inherent properties of the fuel cycle itself that provide the quantity and quality of the fuel below the acceptable threshold in the context of industrial capabilities and economic realities. This idea has also been remarked by Herring et al., [1]. Galperin et al., [8] also identified the requirements for evaluating the fissile material weapon quality by considering three properties: i. Critical mass. A critical mass is different for different isotopic compositions of plutonium and uranium; ii. Weapon yield degradation due to preinitiation caused by spontaneous fission neutrons; and iii. Weapon stability degradation by heat emission. Based on these requirements they [8] provide an approximate comparison of the critical mass for different materials is presented in Table-5. It is clearly demonstrated that a relatively small critical mass is achieved with any plutonium composition, and the RTR-Pu requires 20 to 50 percent more material compared with the weapon-grade material. Table 5 Critical mass for different plutonium compositions. Pu source Critical mass (kg) Weapon grade 4.3 PWR grade 5.5 RTR-seed 5.9 RTR-blanket

33 It was reported that in the LWBR design which is currently being developed in a more deliberately proliferation-resistant way; the central seed region of each fuel assembly will have uranium enriched to 20% 235 U [7, 11]. The blanket will be thorium with some 238 U, which means that any uranium chemically separated from it (for the 233 U) is not useable for weapons. Spent blanket fuel also contains 232 U, which decays rapidly and has very gamma-active daughters. Plutonium produced in the seed will have a high proportion of 238 Pu, generating a lot of heat and making it even more unsuitable for weapons than normal reactor-grade plutonium [11]. Galperin, et al [8, 17] in describing the value of thorium fuel in this respect further asserted that, the probability that an explosive device, constructed from RTR-Pu, will deliver a nominal yield is small (seed) to negligible (blanket), and a probability of a fizzle yield is relatively high. Thus, they [8] conclude, it is shown that the RTR-Pu will produce an unreliable weapon (see table-6 below). Table 6 Probability of an indicated yields [8] Yield Super grade Weapon PWR grade RTR-seed RTR-blanket (Trinity) grade Pu Pu grade Pu grade Pu Nominal fizzle An additional barrier for a possible diversion of a reactor grade material is the heat emitted by its isotopes [8]. The author [8] explained that the thermal power produces an increase of the temperature of a device which makes it unusable for weapons. They remarked that the total heat produced by the RTR-seed and RTR-blanket plutonium is much higher than that produced by the PWR grade plutonium (see table 9, appendix II). Further heat removal measures that may be taken can reasonably impair the stability of the device. Therefore, the RTR fuel cycle which is being developed in a more deliberately proliferation-resistant way, seems technically effective. It is a good supplement that can support the administrative and political measures. In principle, Galperin et al., [8] further commented, all uranium isotopes may be chemically separated from blanket spent fuel and further enriched by standard industrial methods. However, the hard-gamma emissions from decay progenies; and the isotopic composition of uranium are barriers to the diversion. They [8] concluded that diversion route is much more difficult and expensive than simple enrichment of readily available natural uranium. Therefore, in cases where 24

34 enrichment facilities are available, 233 U created in the RTR-blanket does not contribute to the proliferation potential of the fuel cycle. Herring et al., [1] also remarked that the spontaneous neutron generation rate and the amount of decay heat produced by the various plutonium isotopes determine the desirability of a particular mixture of isotopes for use in a clandestine weapon [1]. The properties of the dominant plutonium and americium isotopes are shown in table 10 (plutonium grades and composition are also given in tables - 11 and 12 Appendix II). Besides, from figures in tables - 10 and 12 (Appendix II), the decay heat due to the higher presence of 238 Pu is much higher in thorium-based fuel as compared to uranium oxide that it may be sufficient to melt or damage the other materials (such as RDX high-explosives as their melting temperature limit is ~ C) used in constructing a weapon; Therefore, there is a risk and limitation of producing weapons mainly due to pre-detonation [1]. They [1] further explained that another consideration is the heating of the high explosive surrounding the separated Pu. The high Pu content results in a heat generation of about W for the mixed ThO 2 -UO 2 fuels, compared with about 80 W for the conventional reactor fuel and less than 11 W for the two weapons grades. Using thermal conductivity range of W/mK Table 13 (Appendix II) shows that peak temperatures at the plutonium-high explosive interface are above melting damage point for the high explosives Spent fuel processing, radiotoxicity and decay heat Chemical reprocessing aims at recovering the fissile value of the spent nuclear fuel while removing the wastes, i.e. neutron absorbing isotopes [8]. The uranium and plutonium isotopes found in the spent fuel of a standard LWR fuel cycle present a considerable economic value in terms of natural uranium and separative work [8]. However, spent fuel reprocessing was abandoned in some countries (e.g. USA) as a result of drastic increase in cost of reprocessing due to changing regulatory requirement, public concern to proliferation and other factors [8]. Nevertheless, a closed fuel cycle is being implemented in several countries (e.g. Japan, France, Switzerland) by utilization of mixed-oxide (MOX) fuel containing separated plutonium [8]. These countries base their approach on the economic value of the fissile material contained in spent fuel and by arguing that permanent disposal of spent fuel leaves a (distant) possibility that fissile material could be eventually recovered from a repository and diverted to weapons uses. 25

35 However, contrary to the above view, and in justifying the rationale for not reprocessing the RTR spent fuel Galperin et al., [8] explained that, the relatively small amount of fissile content yield in the RTR fuel cycle from reprocessing may not be attractive [8]. In addition, it also contains large amounts of poison uranium isotopes, such as 234 U, 236 U and mainly 238 U. Presence of a sizable amount of 232 U with high-energy gamma emission would necessitate remote separation and refabrication processing, thereby increasing the overall cost of a reprocessed fuel and reducing the economic incentive for closing the RTR-blanket fuel cycle [8]. They [8] further explained that the amount of plutonium contained in the RTR spent fuel stockpile is much smaller and of a lesser neutronic quality, as shown above. Thus, separation of the uranium component only from the relatively small total amount of the RTR-seed seems economically unjustifiable. In summary, they [8] conclude that, the amount and the isotopic composition of the discharged RTR fuel stockpile makes the reprocessing option of the fuel cycle less attractive from the economic point of view than for a corresponding PWR fuel cycle [8]. The RTR spent fuel can be readily stored and disposed using technology applicable to conventional LWR fuel. This includes considerations of radiological, thermal and toxicity issues [8]. Unlike the RTR fuel cycle other thorium spent fuel can be reprocessed; and it was commented that at least there is a Thorex process that can be used for this purpose [5, 19]. In fuel reprocessing of HEU-Th fuel cycle intended to burn weapon-grade HEU, Vapirev et al., [19] remarked that the technology of reprocessing U-Th fuel is less developed than for U-Pu fuels. However, after justifying the extraction of the minor Pu quantities in the HEU-Th spent fuel as needless; they commented the Thorex process is adequate. Nonetheless, if MEU-Th fuel is used (MEU- 20% enrichment of 235 U) then further improvement in the process is needed for Pu extraction. They also reported that There is a lot of experience in the ITREC prototype plant in Italy and also in the USA, where 870 t of thorium fuel were reprocessed in a commercial facility and 1.4 t of purified 233 U were extracted. Sundararajan et al., [14] also remarked that occupational exposures and environmental releases in the reprocessing stage in thorium fuel cycle are likely to be no different from those in uranium fuel cycle. They [14] further explained that fabrication of thorium subassemblies does not pose any serious radiological problem. However, one of the main drawbacks of thorium 233 U fuel cycle is the presence of hard gamma emitters (2.5 MeV) among the progeny of 232 U which is always present with 233 U. This necessitates shielded and remote handling facility for manufacture of 233 U based fuels. They [14] claim, sol-gel process which uses liquids or free flowing solids for fuel manufacture is ideally suited for remote 26

36 manufacturing facilities. They commented that, as the dose rate on recovered 233 U increases rapidly with time, as shown in Table 7, the fuel fabrication operation should be taken up with out much delay after the recovery of 233 U. They [14] reported that for preparation of ThO 2 -PuO 2 fuel pellets, Bhabha Atomic Research Centre (BARC) has established the sol-gel pelletisation process which minimizes the radiotoxic dust hazard associated with the powder pellet route. Table 7 Dose rate (mgy/h) from 1 kg of 233 U at a distance of 30 cm [14] Aging Time 232 U content in 233 U (days) 100 ppm 1000 ppm

37 CHAPTER 4 Viability of thorium-based nuclear reactors Though thorium fuel is several years away from deploying as commercial power plant, the fuel holds promise of being more efficient, proliferation-resistant, and producing a smaller volume of spent fuel than traditional uranium-based reactor fuel [8]. As many authors commented, there are some challenges that should be sorted out to fully realize the fuel cycle. This chapter mainly evaluate the advantages and challenges exist in utilizing thorium-based fuel systems; and discuss thorium s role in sustaining nuclear energy to the coming centuries. 4.1 Advantages and disadvantages of thorium-based fuel cycles It was observed that thorium-based fuel cycles have many significant advantages over the conventional fuel cycles. Challenges has also been identified that require further improvements; however, considering the little effort given to thorium fuel cycle comparing the opportune U/Pu fuel cycle; these drawbacks could be sorted out if sufficient attention is given with the experience already in hand at different establishments. The feasibility and experience gained in integrating the thorium-based fuel cycles in all existing type of reactors including the required research and development before deployment have been described by the IAEA technical document [4]. It was claimed in the document [4] that, Thorium cycles are feasible in all existing thermal and fast reactors, e.g. LWRs (including WWERs espelially WWER-T), PHWRs, HTGRs, MSBRs and LMFBRs and in ADS. In the short term, it should be possible to incorporate the thorium fuel cycle in some of the above existing reactors without major modifications in the engineered systems, reactor control and the reactivity devices. However, for the innovative reactors and fuel cycles, a lot of reactor physics studies and other technological developments would be required before these could be implemented. In the RTR fuel concept Galperin et al., [8] also reported that replacing a standard (U) fuel by the RTR fuel may have the advantages of: significant of reduction, or if possible, elimination of the fuel cycle proliferation potential; reduction of the spent fuel storage/disposal requirements; and fuel cycle cost saving. 28

38 In referring the two international projects, the Innovative Nuclear Reactors and Fuel Cycles Programme (INPRO) initiated by the IAEA and the US-led Generation IV International Forum (GIF) aiming for future nuclear systems, the IAEA document [4] also reported that thorium fuel cycles is one of the candidates. The IAEA [4] has also identified and summarized in more concise way the benefits and challenges of thorium fuel cycles (below). Most of these advantages and drawbacks were also properly recognized and reported by many authors [5, 6, 21]: 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 of cross-section for thermal-neutrons of 232 Th (7.4 barns). Hence, a higher conversion (to 233 U) is possible with 232 Th than with 238 U (to 239 Pu). Thus, thorium is a better fertile material than 238 U in thermal reactors but thorium is inferior to depleted uranium as a fertile material in fast reactors. For the fissile 233 U nuclei, the number of neutrons librated per neutron absorbed (represented as?) is greater than 2.0 over a wide range of thermal neutron spectrum, unlike 235 U and 239 Pu. Thus, contrary to 238 U 239 Pu cycle in which breeding can be obtained only with fast neutron spectra, the 232 Th 233 U 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 ThO 2 based fuels are one order of magnitude lower than that of UO 2. ThO 2 has favourable thermophysical properties because of the higher thermal conductivity and lower co-efficient of thermal expansion compared to UO 2. Thus ThO 2 based fuels are expected to have better in-pile performance than that of UO 2 and UO 2 -based mixed oxide. ThO 2 is relatively inert and does not oxidize unlike UO 2, which oxidizes easily to U 3 O 8, and UO 3. Hence, long term interim storage and permanent disposal in 29

39 repository of spent ThO 2 -based fuel are simpler without the problem of oxidation. Th based fuels and fuel cycles have intrinsic proliferation-resistance due to the formation of 232 U via (n, 2n) reactions with 232 Th, 233 Pa and 233 U. The half-life of 232 U is only 73.6 years and the daughter products have very short half-life and some like 212 Bi and 208 Tl emit strong gamma radiations. From the same consideration, 232 U 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)O 2 fuel is more attractive, as compared to (U, Pu)O 2, since plutonium is not bred in the former and the 232 U formed after the once-through cycle in the spent fuel ensures proliferation-resistance. In 232 Th 233 U fuel cycle, much lesser quantity of plutonium and long-lived Minor Actnides (MA: Np, Am and Cm) are formed as compared to the 238 U 239 Pu fuel cycle, thereby minimizing the radiotoxicity associated in spent fuel. However, in the back end of 232 Th 233 U fuel cycle, there are other radionuclides such as 231 Pa, 229 Th and 230 U, which may have long term radiological impact. Challenges, The melting point of ThO 2 ( C) is much higher compared to that of UO 2 ( C). Hence, a much higher sintering temperature (> C) is required to produce high density ThO 2 and ThO 2 -based mixed oxide fuels. Admixing of sintering aid (CaO, MgO, Nb 2 O 5, etc) is required for achieving the desired pellet density at lower temperature. ThO 2 and ThO 2 -based mixed oxide fuels are relatively inert and, unlike UO 2 and (U,Pu)O 2 fuels, do not dissolve easily in concentrated nitric acid. Addition of small quantities of HF in concentrated HNO 3 is essential which cause corrosion of stainless steel equipment and pipings in reprocessing plants. The corrosion problem is mitigated with addition of aluminum nitrate. Boiling THOREX solution {13 M HNO M HF+0.1 M Al(NO 3 ) 3 } at ~393 K and long dissolution period are required for ThO 2 -based fuels. The irradiated Th or Th-based fuels contain significant amount of 232 U, which has a half-life of only 73.6 years and is associated with strong gamma emitting daughter 30

40 products, 212 Bi and 208 Tl with very short half-life. As a result, there is significant buildup of radiation dose with storage of spent Th-based fuel or separated 233 U, 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 232 Th to 233 U, 233 Pa is formed as an intermediate, which has a relatively longer half-life (~27 days) as compared to 239 Np (2.35 days) in the uranium fuel cycle thereby requiring longer cooling time of at least one year for completing the decay of 233 Pa to 233 U. 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 233 U and thorium. The three stream process of separation of uranium, plutonium and thorium from spent (Th,Pu)O 2 fuel, though viable, is yet to be developed. The database and experience of thorium fuels and thorium fuel cycles are very limited, as compared to UO 2 and (U,Pu)O 2 fuels, and need to be augmented before large investments are made for commercial utilization of thorium fuels and fuel cycles. Herring et al., [1] also analyzed and compared the cost of a thorium-based fuel cycle and uranium-oxide fuel. The results of this economic comparisons among two uranium dioxide fuels and two mixed ThO 2 - UO 2 fuels irradiated to 72 and 87 MW day/kg are shown in Table 14 and 15 (Appendix III) [1]. It was noted that a higher price for uranium and a slight decrease in the thorium and fuel fabrication prices result in about a 9% cost advantage for the mixed ThO 2 - UO 2 fuels as compared with the high burnup all-uranium fuel. They [1] explained further, the cost of mixed ThO 2 - UO 2 fuel should be relatively insensitive to the price of thorium, since no enrichment is required (or possible since 232 Th is the only isotope). Therefore, no enrichment tails are generated and only kg of natural thorium is needed per kg of fuel. Unak [6] also remarked the costs of thorium fuel cycle that will probably be reduced by about 20-30% than uranium fuel cycle; especially eliminating the isotope enrichment processes; such as in the RTR concept which do not require fuel reprocessing. He [6] also identified drawbacks of thorium fuel as, among others, the need of initial fissile material like 235 U or 239 Pu to produce 233 U and the poor dissolution behaviour of thorium oxide which is not easy as that of uranium oxide and this needs more efforts for reprocessing of spent fuels. 31

41 Aiming to reduce plant-operating costs and price of electricity produced by nuclear power plants in the US, Herring et al., [1] study also show that with improved long fuel burnup, which would increase plant capacity factors by about 5% can reduce a production cost to 2 cents per kw-h which make nuclear energy more competitive. Besides, most worker exposure and low level waste generation at commercial nuclear plants occurs during refuelling, longer refuelling cycles will also reduce worker exposures to radiation and the amount of low-level waste generated [1]. Furthermore, Lung and Gremm [5] assert that 233 U maintains its good neutronic properties with higher temperatures, better than either 235 U or 239 Pu for most neutron energies; therefore these led to recommend the thorium fuel cycle for high temperature reactors. In addition, they remarked that thorium oxide, uranium oxide and plutonium oxide have similar physical characteristics and can cyncristallize in the centered cubic form. They [5] reported that This property is most important for the manufacture and stability of hybrid oxide fuels and no doubt permits the manufacture of very high burnup fuels. 4.2 Thorium and Sustainability of Nuclear Energy Sustainability considerations mainly underline a view of a system from aspects of such as, environmental soundness, addressing social concern and the net benefits that it provides to people which include economical viability. It was noted earlier that thorium can be a good supplement to the uranium fuel cycles, which is limited in abundance, to sustain nuclear energy in the coming years. Thorium fuel cycle is an attractive source of nuclear energy that generates relatively low radiotoxicity and long lived minor actinide elements (MAs: Np, Am, Cm) than the U/Pu fuel cycles [21]. In addition, thorium-based reactors could be used to incinerate accumulated weapons grade plutonium (WPu) or civilian plutonium. Besides, thorium fuel is more proliferation-resistant than the uranium fuel cycles, mainly due to the presence of high energy gamma-rays in the system and heat production in the separated fissile materials. These factors greatly address some of environmental and public concerns though much is still expected. Calculations made by Herring et al., [1] justify the fuel cycle economical advantages; and show that the fuel costs for a mixed ThO 2 -UO 2 core and for an all uranium core burned to 72 MW day/kg are about the same and about 10% higher than for an all-uranium core burned to only 45 MW day/kg. However, these analyses have not included the economic advantages associated with increased plant capacity factors and reduced outage costs, due to the longer 32

42 fuel cycles. The costs for a mixed ThO 2 -UO 2 core burned to 87 MW day/kg are a few percent less than the costs for the 72 MW day/kg cores, but not significantly different [1]. Galperin, et al [8] also declared the fuel cycle cost savings of RTR approximately is 20%. The amount of natural uranium per cycle is about 140 Mt compared with 170 Mt of a standard PWR, and the number of fuel rods fabricated is 11,000 compared with 15,000 rods per PWR annual reload. The fabrication cost of a metal alloy fuel rod (RTR-seed fuel), produced by an extrusion process, is significantly lower than that of a PWR oxide rod. These three components of the front-end, as well as back-end savings in spent fuel storage expenses, result in a 20 to 25 percent reduction in an overall fuel cycle cost. This estimate is supported by a detailed fuel cost calculation [8]. Lung and Gremm [5] also commented that in the future, 235 U will become precious. They remarked that thorium is as fertile a material as 238 U will also be used in the long-term. Besides, referring Furukawa and Bowman ideas, they [5] remarked, the breeders of the future may be very different from today s fast reactors, for example being enhanced by accelerators to obtain fissile material more quickly. They [5] further stated that accelerator-driven reactors must be regarded as being complementary to fast breeders; a comparison might show which will be the breeder of the future; maybe both have a chance. They concluded that For the many reasons explained above, thorium has not thus far been able to compete on a par with uranium, which occupies a privileged niche. The time has come to have another hard look at what was perhaps too quickly set aside twenty years ago and start anew with fresh ideas. The accelerator-driven reactor is just one of them. Thorium reserves are estimated to be about 4 times those of uranium. By contrast, thorium is not soluble in water and its extraction from sea water is not being considered, contrary to uranium. Here, It should be stressed that breeder reactor technology being very economical in its use of the fuel, even low content ore could be worked profitably, ensuring that fuel would remain available over several thousand years both for uranium and thorium-based breeder reactors [13]. The fact thorium reserves are large is thus not important issue. Particularly, from strategic point of view those countries with sufficient reserves, like India and Brazil should advance further to this aim along with international cooperation. 33

43 CHAPTER 5 Thorium fuel and emerging advanced reactor concepts Nuclear energy remains a good possible technology to produce large amounts of energy in better conditions of safety and durability without large air pollutant gas (like CO 2 ) emissions. Several innovative concepts are now under study in order not only to produce electricity with high conversion efficiency but also to supply high temperature heat for combined cycle, such as hydrogen production and water desalination. This chapter introduces and discusses nuclear reactors applications; and advanced nuclear reactors and fuel cycles in order to meet the global energy needs in the 21 st century and beyond. 5.1 Nuclear applications and potentials of thorium-based reactors Nuclear reactors applications can generally be categorized into three, as current power generation, which accounts 16% [4] of world electric production; current heat applications, which are combined heat and power (CHP) applications that are low temperature usages such as district heating, desalination; and additional future high temperature heat applications. There are also experimental reactors for material testing, isotope production and training purposes. Besides, reactor types for space and naval propulsion applications exist. Classification of reactors by generation also has been made; and figure 8 below shows the overlapping life cycles of different reactor generations (I-IV). Figure 8 The overlapping life cycles of the different reactor generations [2] 34