Thoria and inert matrix fuels for a sustainable nuclear power

Progress in Nuclear Energy 50 (2008) 944–953

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Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene

Thoria and inert matrix fuels for a sustainable nuclear power C. Lombardi a, L. Luzzi a, E. Padovani a, *, F. Vettraino b a b

Department of Nuclear Engineering, Politecnico di Milano, via Ponzio 34/3, 20133 Milano, Italy Nuclear Fission Division, ENEA, via Martiri di Monte Sole 4, 40129 Bologna, Italy

a r t i c l e i n f o

a b s t r a c t

Dedicated to Professor Bruno Montagnini, University of Pisa.

Nuclear power to be sustainable requires the fulfilling of peculiar constraints, which include addressing the proliferation risk. One possible route for sustainability is that to adopt a fuel cycle based on thorium. However, comparison with uranium cycle indicates that thorium cycle utilization is premature. Instead, a promising short-term option is the use of inert matrix fuels, possibly containing thoria, in a oncethrough cycle. Irradiation tests performed in the Halden reactor show encouraging behaviour under irradiation. Furthermore, these fuels are very well suited for a direct disposal in a geological repository. Ó 2008 Elsevier Ltd. All rights reserved.

Keywords: Inert matrix fuel Proliferation Sustainable energy Thorium

1. Introduction The concept of sustainable development was brought into common use by the Brundtland Commission (WCED, 1987), who defined it as the development that meets the needs of the present generation without compromising the needs of future generations. Other proposed definitions stem from that one, but putting more emphasis on the conservation of the natural environment (NSESD, 1992) or on the quality of life (Scott, 2004), while others are focused on poverty alleviation (World Bank, 2002). Generally, sustainable development is thought to have three facets. The economical facet deals with costs and resource depletion; the environmental facet is related to the preservation of so-called global life support services; the social facet involves social–political issues, being, for what the energy production is concerned, mainly related to acceptability by political institutions and public opinion. In the last years, the environmental facet of energy exploitation has been focused on CO2 emissions. Authoritative studies, as the Intergovernmental Panel on Climate Change (IPCC) reports, point out the possible dramatic climate changes due to the growing rate of greenhouse gases emission. On the other hand, the availability of energy, and above all electricity, is one of the key factors needed to the underdeveloped countries to be freed from poverty and to developed countries to support economy and social services. Mankind strongly relies on energy, then it is sometimes claimed that the most hazardous energy is the one not dispatched when needed. Nuclear energy had been early recognized by scientists and technicians as environment friendly (see e.g. http://www.ecolo. org) for its low pollution, in particular its low specific CO2 emission, being CO2 mainly produced in the front-end of the fuel cycle and in * Corresponding author. Tel.: þ39 02 23996398; fax: þ39 02 23996309. E-mail address: [email protected] (E. Padovani). 0149-1970/$ – see front matter Ó 2008 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2008.03.006

the construction of power stations. Presently, this attitude is also shared by some influent historical environmentalists, who are concerned of the urgency to adopt low carbon emitting energies, because we have no time for visionary energy sources (Lovelock, 2006). Also from the ONU-IPCC AR4 Synthesis Report (ONU-IPCC, 2007) it may be argued that nuclear power is today an effective greenhouse gas mitigation option, and that a robust mix of energy sources, including nuclear, will almost certainly be required to meet the growing demand for energy services. However, nuclear power to be sustainable requires the fulfilling of conditions which sensibly differ from the ones to be applied to conventional sources. Besides the usual criteria involving cost, safety, resource depletion and impact of waste, in the case of nuclear power the public acceptance and proliferation concerns assume a special relevance. Generally, the public opinion is less confident on nuclear safety than on safety of other resources, both fossil and renewable; this higher perception of risk has as a consequence that more stringent safety criteria are applied to nuclear power. These criteria are indeed satisfied by the present generation of nuclear reactors. The public opinion is now more concerned about the legacy to future generations of long-lived radioactive wastes. Mainly for this reason, the adoption of a closed fuel cycle, which implies the deployment of breeder reactors, is the main path to follow in the development of the next generation’s nuclear power. Moreover, the issue of the potential chance of nuclear weapons proliferation is likely to be the hardest obstacle to a further deployment of peaceful nuclear power, able to make acceptable or not its expansion at a global scale. In the present paper, Section 2 is devoted to a synthetic review on thorium utilization. Thorium cycle studies have a long story, since the beginning of the nuclear era, without resulting into extensive applications, but its revival was pushed up by a completely new proposal based on a uranium–thorium closed fuel cycle

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applied to a fast reactor, connected to innovations in the reprocessing and plant scheme, still under study (Rubbia et al., 1995a,b). In Section 3 a brief review of the thoria fuels characteristics is reported. Thorium fuels were and are sporadically used, especially in India; however, in spite of their promising features (Lung, 1997; ¨ nak, 2000; IAEA-TECDOC-1155, 2000; Lung and Gremm, 1998; U IAEA-TECDOC-1319, 2002; IAEA-TECDOC-1450, 2005), the hope of their rapid utilization appears premature. In Section 4 the issue of radioactive waste is addressed, while Section 5 is devoted to briefly discuss the proliferation issue, and some promising ways for effectively hampering or preventing the misuse of plutonium are described. These two issues probably represent the most relevant ones for the perspectives of the nuclear energy source. Section 6 deals with fuels which do not produce plutonium; on the contrary, they are fit to effectively burn it in conventional water reactors and they can be conveniently sent to disposal without the need of reprocessing. Suitable to be implemented in the short term, this solution was proposed some years ago (Cerrai and Lombardi, 1992) and is still under development (Journal of Nuclear Materials, 1999; Progress in Nuclear Energy, 2001; Journal of Nuclear Materials, 2003). In Section 7 the conclusions of the paper are drawn. In the following, with open cycle or once-through cycle is intended that the burned fuel will be disposed without reprocessing. With closed cycle is intended an indefinitely repeated chain of fuel manufacturing, burning and reprocessing, in which only fertile materials need to be added. If the reactors are not able to breed enough fissile, a similar fuel chain can be adopted, with the aid of new fissile added at each recycle (fissile topping); in this case we can speak of a semi-closed cycle. Nowadays, it is generally intended that in a closed cycle all actinides are recycled. About the fuel, UO2, MOX and TOX are referred to, the last being a mixture of thoria and plutonia and/or urania. 2. Comparison between the thorium–uranium and uranium– plutonium cycles 2.1. Availability of uranium and thorium Concerning the conventional uranium resources, official documents refer to values in the range 4–5 millions of tons at <130 $/kg, which would be exhausted in about 70–80 years at the present rate of consumption. However, considering also the speculative and/or unconventional resources, we can state that the uranium resources are much more abundant (OECD/NEA and IAEA, 2005). The thorium resources could be even larger. In the case where breeder reactors are adopted, the resources of both the elements become almost inexhaustible. 2.2. Conversion The principal reactions induced by thermal neutrons, starting from 232Th (thorium–uranium cycle) and from 238U (uranium– plutonium cycle), are reported in Fig. 1. The similarities are remarkable. The main parameters, which influence the breeding performance of the thorium–uranium and uranium–plutonium cycles, are the fertility factor h (mean number of neutrons produced per neutron absorbed) of the fissile and the fission cross-section of the fertile. (1) Approximately, we can state that 233U’s fertility factor is greater than 239Pu’s one for neutrons energy up to a few tens of keV, while the opposite occurs for neutrons of greater energy (Fig. 2). (2) The fission cross-section of 232Th is smaller than the 238U’s one, furthermore the fission threshold energy of the former is greater.

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Item (1) implies that at thermal energies the thorium–uranium cycle is superior with respect to uranium–plutonium cycle; items (1) and (2) imply that in the fast region the opposite is true. Further arguments concerning the conversion factor and core feasibility are  For thorium, item (2) is a serious drawback for the conversion factor, but it is an advantage for what the coolant void coefficient is concerned, both in water and fast reactors. For instance, in a published article (Lombardi et al., 2001) it is shown that using thorium as fertile instead of 238U in a once-through cycle in a PWR, the fuel pins can be loaded with a greater amount of reactor grade plutonium, then reaching a much higher burnup, without incurring in positive void reactivity coefficients.  The build-up of other isotopes is also to be taken into account, in particular of the fissile ones 235U and 241Pu. The latter exhibits particularly large values of both the fertility factor and the cross-section, but it is partially lost due to decay into 241Am.  The relatively long half life of 233Pa leads to two adverse effects: (i) when the reactor is in operation a small fraction of this isotope undergoes radiative capture, with a consequent waste of 233U; (ii) during reactor shut-down a small 233U build-up occurs. Both these effects are of modest magnitude; however, they are to be taken into account. Synthetically:  In reactors with neutron spectrum characterized by a high thermal component (i.e. well-moderated graphite or heavy water reactors) the thorium–uranium cycle allows to obtain conversion values greater than the uranium–plutonium cycle’s ones. In particular, it is possible to attain breeding in CANDU reactors. The Shippingport reactor experiment proved that also for a PWR it is possible to reach breeding conditions, but at the cost of great complications in the core design and management.  In fast reactors the uranium–plutonium cycle allows to obtain breeding gains definitely higher than the thorium–uranium cycle’s ones.  To obtain a high conversion factor in light water reactors, it is necessary to lower the moderation ratio as much as possible, so achieving an Advanced High Conversion Water Reactor, like the Japanese Project RMWR (Iwamura et al., 2006; Suzuki et al., 2004). The resulting neutron spectrum lacks the thermal component, then the uranium–plutonium cycle is advantaged but, to keep negative the void coefficient, a complex geometry of the core, shaped like a multi-layer sandwich, is needed. A distinctive way to improve the conversion factor consists in taking advantage of symbiotic systems. For instance, in a book published by Ronen (1990) a system of two reactors derived from PWRs is proposed. The first reactor is fuelled with thorium–plutonium and produces 233U, the second one is fuelled with 238U–233U and produces plutonium. Clearly, enlarging the possible choices so as to include also fast reactors, higher conversion factors can be attained. For instance, a fast reactor could be fuelled with thorium, a little 238U and plutonium, to produce denatured 233U, while a PWR could be fuelled with 238U–233U to produce plutonium. By this way in both reactors the loaded fissile has the higher fertility coefficient; this partially compensates the penalty on conversion due to the presence of thorium in one of them. 2.3. TOX fuel cycles Thorium fuel cycles have a long history (IAEA, 1970). For instance, Indian Point I, one of the first commercial nuclear power

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232

233

Th+n

FERTILE

β

Pa

238

239

U+n

β

Np

27.4 days

2.3 days

fission

fission 233

90

U+n

FISSILE

10

capture

234

239

Pu+n 35

capture

U+n

FERTILE

240

U+n

FISSILE

241

65

Pu+n

fission

fission 235

80

20

capture

236

U+n

Pu+n

capture

PARASITE

237

Np Chemically separable

242

75

25

Pu+n

243

Am Chemically separable

Fig. 1. Analogy in thorium and uranium fertilization.

plants, and Elk River were designed and operated during their first cycle with ThO2–UO2 fuels, and many ThO2–UO2 fuel bundles were irradiated during the LWBR (Light Water Breeder Reactor) program, conducted in the 1960s and 1970s at Shippingport (Goldberg et al., 1979). These cycles were directed toward the production, reprocessing, and reuse of 233U through reactors having mainly seed-and-blanket configurations, utilizing highly enriched 235U for start-up, and achieving burnups of 30 MWd/kg. At various times since, interest in thorium cycles has been renewed for various applications, as attested by several studies carried out during the International Nuclear Fuel Cycle Evaluation Project (IAEA, 1980) and the Non-proliferation Alternatives Systems Assessment program (DOE, 1980), which was related to the use of thorium as an anti-proliferation fuel. Review of pertinent literature (Weaver and Herring, 2003; MacDonald and Lee, 2004) brings to light three shortcomings relative to present needs: (i) the previous studies presumed LWR fuel reprocessing and recycle; (ii) start-up fissile needs were provided using fully enriched (w93%) 235U fuel; and (iii) state-of-the-art fuel burnup for PWRs was then only 25–30 MWd/kg, about half today’s best capability, and only 25–30% of what is potentially possible. Circumstances have changed since then, and the nuclear industry in most countries is now constrained by a limitation to once-through fuelling, the use of 20 wt% 235U,1 and the ability to seriously

1 This value is the maximum allowed to eliminate the possibility to build a weapon; however, no commercial fuel factory in the world is licensed to handle enrichments higher than 5%.

consider in uranium-fuelled LWRs a discharge burnup approaching 70 MWd/kg, with further increases in prospect. More lately, (Th,U)O2 fuels have attracted again attention, stirred up by the need for longer fuel cycles, higher burnup and performance, and improved waste form characteristics (Zhao et al., 1999; Fourest et al., 2000; Herring et al., 2001; Galperin et al., 2002; Loewen et al., 2002). It is worth mentioning the International Project on ‘‘Advanced Nuclear Fuels/Fuel Cycles’’ sponsored by the U.S. Department of Energy in the framework of the Nuclear Energy Research Initiative (NERI Project 00-014, FY 2000), to which a special issue of Nuclear Technology (vol. 147, July 2004) has been dedicated. This Project evaluated the efficacy of the ThO2–UO2 once-through fuel cycle in current LWRs with high fuel burnups (MacDonald, 2000, 2001; DOE, 2003; MacDonald and Lee, 2004), and namely focused on homogeneously mixed (Saglam et al., 2004; Joo et al., 2004) and micro-heterogeneously mixed (Shwageraus et al., 2004a) TOX fuel cycles, which rely on in situ use of the bred-in 233 U. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels have not shown any economic advantage over UO2 fuel, when current fuel management strategies are used; in fact, the main results of this Project indicate the following (Lahoda, 2004): (i) in all cases, the (Th,U)O2 fuel costs more than a comparable all-uranium fuel (the percentage increase is at least 25% for a ThO2–UO2 ratio of 75/25, but it could be up to 64% if this ratio is changed to 70/30); (ii) the (Th,U)O2 fuel uses more uranium resources than the uranium-only fuel, because of the higher enrichment required. Since the economics of (Th,U)O2 cores was not as promising as originally expected, work in NERI Project turned to looking at using thorium as a strategy for burning

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3.6 3.4

233U

3.2

235U

3

239Pu

2.8 2.6 2.4 2.2 2 1.8 1.6 1.4 1.2 1 10−2

10−1

100

101

102

103

104

105

106

107

eV Fig. 2. The fertility factor for the main fissile isotopes (elaboration of ENDF/B VI data).

unwanted reactor- or weapons-grade plutonium as well as longlived minor actinides, in order to reduce the radiotoxicity in spent nuclear fuel (Weaver and Herring, 2003; Shwageraus et al., 2004b; Dziadosz et al., 2004; Herring et al., 2004). The feasibility of using TOX fuel in closed or semi-closed cycles can be compared with the MOX one. A semi-closed cycle using MOX fuel in standard LWRs can hardly be sustained. The quality (fissile fraction) of plutonium quickly degrades generation after generation, and its reuse is hampered by the emerging of positive void coefficients. This drawback can be avoided or reduced by mixing the recycled plutonium with enriched uranium and/or topping it with high quality plutonium (e.g. weapon grade). However, the full recycle of all the minor actinides cannot be pursued in any case, as they become more and more radioactive. In particular, the build-up of little amounts of 252Cf makes the fabrication of new fuel harder and harder, due to the very strong neutrons emission. The substitution of MOX fuel with TOX reduces these drawbacks; furthermore, thanks to the contribution of 233U, TOX exhibits a better conversion factor in thermal reactors. In a word, TOX fuel is globally better than MOX. Unfortunately, the recycling of TOX fuels is today a difficult task. Nowadays, there is little advantage in reprocessing the UO2 fuel, and no advantage in reprocessing MOX (poor plutonium quality) or (Th,Pu)O2 (reprocessing not yet assessed at industrial scale). In the future, after industrial facility for thorium fuels will come into operation, we think that both uranium-based and thorium-based fuels will be adopted. Indeed, it appears that fast reactors will be needed to exploit all the benefits of closed cycles. Today, TOX fuel can be adopted to take advantage of its superior performance in burning excess plutonium in a single passage in thermal reactors, particularly in HTGRs.

3. Basic features of TOX fuels There are several differences between the properties of ThO2 and UO2 fuels that may result in differences in their performance during in-service conditions (Belle and Berman, 1984; Loewen et al., 2000, 2001). Compared to UO2, the main ThO2 fuel properties consist in: (a) a higher thermal conductivity, (b) a much higher (w400–500  C) melting temperature, (c) a lower coefficient of thermal expansion, (d) more fission gas production for 233U than 235 U fission, (e) a lower value of the diffusion coefficient for fission gases, and (f) a significantly less (w10%) theoretical density (TD). This last feature, together with the lesser heat capacity of ThO2 fuel relative to UO2 fuel, results in a lower stored thermal energy per unit volume in ThO2 fuel.

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The thermal conductivity is one of the most important properties for estimating the behaviour of nuclear fuel during both normal operation and postulated accident conditions. It is well known that the thermal conductivity of ThO2 (Bakker et al., 1997) is higher than that of UO2 (Fink, 2000), namely by w50% over a significant range of temperature (Long et al., 2004). Some information is available in literature for thoria–urania mixtures (Srirama Murti and Mathews, 1991; Lucuta et al., 1994), but more data are still needed to completely characterize the thermal conductivity of (Th,U)O2 fuel pellets. As a rule, in a homogeneous unirradiated mixture of ThO2– UO2, the thermal conductivity is somewhat higher than the thermal conductivity of unirradiated UO2, depending on the temperature and the relative content of the ThO2 (Belle and Berman, 1984); nevertheless, it is worth mentioning that thermal conductivities of (Th0.655U0.345)O2 and (Th0.355U0.645)O2 pellets were found to be lower than that of both pure ThO2 and UO2, such degradation by UO2 substitution being large at low temperatures, but smaller as the temperature increases (Yang et al., 2004). Data for various homogeneous ThO2–UO2 mixtures of 95% theoretical density, which have been irradiated up to a burnup of 30 MWd/kgHM, indicate that the thermal conductivity of ThO2–UO2 fuel is greater than that of 100% UO2 fuel when the fraction of UO2 in the mixture is less than w20% (Long et al., 2004). TOX pellets can be fabricated by ceramic powders processing, i.e. by sintering a mixture of ThO2, UO2 and/or PuO2 powders: unlike single-oxide (UO2) pellets, processing of TOX fuels is generally more complex because it is difficult to get a homogeneous mixture of the different powders by mechanical mixing or milling and, moreover, sintering of powders should result not only in densification, but also in the formation of a solid solution. Anyway, other mixed oxide fuels – e.g., (U,Pu)O2 and (U,Gd)O2 – have been successfully fabricated in a similar way (Assmann and Robin, 1983). The most relevant differences in TOX powder processing, in comparison with UO2 pellets fabrication, are related to (i) the high energy gamma emitting daughter products of 232U and (ii) the much higher melting temperature (3640 K of ThO2 vs. 3120 K of UO2). The first item would require complete remotisation of processes, which would result more expensive. In this regard, impregnation techniques for ThO2–UO2 fuels, which have been recently developed at Bhabha Atomic Research Centre (Basak et al., 2004) appear to be more suitable for the fabrication of highly gamma active 233U bearing mixed oxide fuel pellets, since the major part of fabrication processes can be carried out in an unshielded facility. The second item, which is absolutely beneficial as concerns the fuel performance during in-service and hypothetical accident conditions, implies that the sintering of ThO2 would require higher temperatures than that of UO2. Seeing that it is difficult and uneconomical to raise the sintering temperature of ThO2 above 1800  C, ThO2 or (Th,U)O2 pellet fabrication needs the preparation of powder that can be sintered at w1700  C (Yang et al., 2004). To this purpose, several powder preparation methods have been investigated to increase the sintering activity of pure ThO2 powder (Harada et al., 1962; Pope and Radford, 1974; Anthonysamy et al., 2000), and also sintering agents such as MgO and Nb2O5 have been used to fabricate ThO2 pellets with appropriate density (95% of TD or greater) and strength at desirable sintering temperatures (Balakrishna et al., 1998). In this short overview on the basic features of TOX fuels, other two relevant issues also need to be mentioned, namely: (i) their behaviour during normal, off-normal, and accident conditions, in comparison with that of UO2 fuel, according to the U.S. NRC (Nuclear Regulatory Commission) licensing standards; (ii) the long-term stability of TOX waste. These two issues have been thoroughly studied in the framework of the NERI Project 00-014 (DOE, 2003; MacDonald and Lee, 2004). For details see Nuclear Technology, vol. 147, July 2004. Hereafter, some general

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considerations are outlined with reference to (Th,U)O2 fuels under LWR conditions. As regards the first issue, a broad range of analyses (Long et al., 2004) – by means of widely used LWR calculation tools modified and extended for ThO2-based fuels – has indicated that the thermal, mechanical and chemical performance of homogeneous ThO2–UO2 LWR fuels with respect to safety is generally equal to or better than that of all-uranium fuel, during both normal operation and accident conditions. The irradiation performance of these TOX fuels is better than urania, because of some of their favourable material properties, such as the higher thermal conductivity and melting temperature, and the lower diffusion coefficient for fission gases with respect to UO2 (Long et al., 2002; Lee et al., 2004; Long et al., 2004). The fission gas diffusion coefficients have an exponential dependence on temperature and are among the most basic and important data needed to properly model the fuel rod behaviour in reactor, especially at high burnup. The reason why mixing ThO2 and UO2 reduces the diffusion coefficients is not known; however, the effect has been clearly shown at KAERI (Korea Atomic Energy Research Institute), where several annealing tests on UO2 and ThO2–UO2 specimens (with a ratio of thorium and uranium of 35:65) were performed, after irradiation in the Hanaro reactor: the 133Xe release fraction for each specimen was obtained as a function of temperature (from 1400 to 1600  C) and the conclusion was that polycrystalline (Th–U)O2 has one-order-of magnitude lower 133Xe diffusion coefficients than the polycrystalline UO2 (Kim et al., 2004). Therefore, the fission gas release in a homogeneous mixture of ThO2–UO2 fuel is expected to be lower than the one of 100% UO2 fuel, under the typical conditions of current PWR fuel rods (Belle and Berman, 1984; Long et al., 2004; MacDonald and Lee, 2004). The reasons are (i) the lower diffusion coefficient for fission gases, and (ii) the probable reduced fuel temperature, due to the probable higher thermal conductivity of TOX (UO2 fraction less than w20%). The proposed long-term storage of spent fuels in geologic repositories has given rise to questions regarding the release of uranium and other hazardous radionuclides into the surrounding environment. Results from experiments (Demkowicz et al., 2004) concerning the aqueous dissolution of irradiated thoria–urania LWR fuel pellets in Yucca Mountain well water have demonstrated that (U,Th)O2 exhibits a measurable improvement over UO2, suggesting that urania–thoria spent fuel results in a more stable and insoluble long-term waste form than conventional UO2 fuel for direct disposal in the repository (Sunder and Miller, 2000). As a matter of fact, the dissolution rates of irradiated ThO2–UO2 pellets with compositions ranging from 2.0 to 5.2% UO2 were found to be at least two orders of magnitude lower than reported literature values for pure UO2 (Demkowicz et al., 2004). The formation of a more stable waste form for (Th,U)O2 fuels is connected to the following basic difference between thorium dioxide and uranium dioxide: UO2 is not the highest oxidation state for uranium (it will further oxidize to U3O8 and UO3), and because of the volume changes associated with the further oxidation of UO2, the current LWR spent fuel is not as robust a longterm waste as ThO2, which is the highest form for thorium oxides and is a very stable material that exhibits very low solubility in groundwater – generally <1010 mol/l at 25  C (Fourest et al., 2000). On the other hand, wet reprocessing of (Th,U)O2 fuels is more complex, owing to their more difficult aqueous dissolution; moreover, it requires remote handling and heavy shielding, due to the decay products of 232U and 228Th (Mah, 1983). It is worth recalling that the chemical separation of uranium from thorium is also more demanding than separation of plutonium from uranium, since the former requires larger volumes of processing materials (Wilson, 2002; IAEA, 1970).

4. Radiotoxicity of the waste and impact on the environment Generally, the specific radiotoxic inventory, i.e. the committed dose due to the ingestion of discharged fuel which had produced 1 GW year of electric power, is reported in the literature. Excluding the first period, during which the contribution of medium-lived fission products prevails, the radiotoxic inventory of a typical burned LWR fuel is almost entirely given by the actinides, in particular by plutonium (IAEA-TECDOC-840, 1995; IAEA-TECDOC-916, 1996). It is possible to lower the specific radiotoxic inventory at medium-long time of the wastes put under ground: (i) by ten times in fast reactors, recycling all the actinides in a closed cycle; by w2 times in LWRs, recycling the plutonium only once in MOX fuel (Kloosterman, 1998; Gruppelaar et al., 1998). The latter is the only easy, industrial way of waste recycle. A shift to thorium–uranium cycles (with fissile topping for not-breeding LWRs) might lower the waste’s radiotoxic inventory of both water and fast reactors. The most effective way to reach this goal foresees the use of an advanced thorium–uranium closed cycle coupled to a novel accelerator-assisted fast reactor (Rubbia et al., 1995a). However, there are some open issues about Accelerator Driven Systems (ADSs), namely: the behaviour of materials in contact with liquid lead or lead–bismuth, the required performances of the accelerator, its coupling with the core target, and the fuel reprocessing. Then, it is likely that ADSs will be very complicated, expensive and scarcely reliable, thus unsuited to a widespread energy production (the Energy Amplifier concept). The best chance for the deployment of ADSs will consist in reserving them to the burning of minor actinides in a double-strata strategy (OECD, 2002), to be postponed far in the future. Furthermore, it is worth remembering that the radiotoxic inventory, or potential radiotoxicity, is very different from the dose which is actually imparted. If we imagine that the fuel is buried into a suitable geological repository and we consider the normal evolution scenario (Boussier et al., 2000), the committed dose to the critical group of the population, which turns out to be much lower than the natural background,2 generally, is not due to actinides, but due to some fission products which are able to migrate more easily through the engineered and natural barriers. If we consider the chance, even many thousands of years after the burial, that an accidental human intervention (inadvertent human intrusion scenario) could take back to surface a significant (a drilling core) amount of waste, very high doses are imparted both by actinides and some long-lived fission products. From these arguments we can infer that the recycling of actinides, in particular of plutonium, can greatly decrease the radiotoxic inventory, but has a modest impact on the doses really imparted by the waste when put in the geological repository, because the remaining fission products are more apt to migrate. Indeed, some authors are sceptical on the actual timeliness of recycling the fuel, considering that this option surely allows to avoid a potential radiological risk in the future, but at the expense of running a real risk in the present for the doses imparted, mainly to the workers, during the fuel reprocessing and fabrication; moreover, large amounts of separate plutonium must be handled, then increasing the proliferation risk. The fuel recycling, through the plutonium and uranium recovery, allows a resource saving. In the actual water reactors the saving is modest, about 25%. For what the dose to workers and population is concerned, the dose decrease due to the saving of mining and milling activities is almost vanished by the dose due to the reprocessing. Also from the economic point of view, the saving of fresh uranium and separation work are offset by the additional

2 The assumed waste quantity corresponds to the generation of 165 GWe year by PWRs in once-through cycle.

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reprocessing e fabrication costs. However, we can reasonably assume that the environmental constraints will be more stringent in the future and that the costs and doses related to the recycle option will decrease along with the technology progress. Waste recycling could become a crucial argument for the nuclear option to become desirable to a vast public, concerned by the present heavy footprint of mankind on our planet (Wackernagel and Rees, 1996). Then, we are witness of a schizophrenic technical attitude; if we give credits to those anxieties, we have to admit that the recycle option is beneficial for a better resource exploitation and it allows to lower the environmental impact (principally the one perceived by the public opinion) given by the mining of fresh uranium and by the radiotoxicity of the waste. Presently, there is a new, real technical argument in favour of reprocessing: it permits lowering the thermal load of the waste put in the disposal, therefore the repository capacity can be greatly increased. This is the case of the Yucca Mountain repository (DOE, 2002a), for which a change of mind on the timeliness of adopting a once-through cycle is under way. 4.1. Partial conclusions The present exploitation of nuclear energy, which is mainly carried out through water reactors fed by enriched uranium and operated in once-through cycle or one recycle, is sustainable today under the view-point of safety, economics, resource availability and impact on environment. The public acceptance can be improved if the radiotoxicity of the waste to be put in the disposal will be lowered. Also the impact of the mining and milling activities could give rise in the future to the resistance of the concerned people. Considering that the environmental constraints will become more and more stringent and that it takes tens of years for relevant changes to reactors and fuel cycles to take place, it will be better considering, for future developments and deployments, breeder reactors and closed cycles (Wilson, 2000). Actually, breeding capabilities are present in most of the proposed Generation IV reactors (DOE, 2002b). Indeed, it is worth reminding that fast breeder reactors have been developed in the past, with really impressive investments, in many nations, but technical and economical problems have hampered their deployment. About the thorium, its large-scale exploitation, like the 238U one, was envisaged through a closed cycle in thermal, fast or symbiotic breeder reactors, with the aim to have an almost inexhaustible energy resource. Today, the interest is mainly focused on the lower radiotoxicity of the produced actinides. Alternative to the closed cycle, an indefinite recycle in conventional PWRs of thorium bearing fuel (topped with fresh plutonium) has been proposed as an effective way of completely destroy the topped plutonium and exploit its energy potential (Lombardi et al., 1999a). The preliminary conclusion is that the exploitation of nuclear energy is sustainable in the short and medium term. However, in the future it will be necessary to recycle the waste and to adopt reactors with a higher conversion factor, possibly fast breeders. 5. The proliferation issue A concern linked to the present nuclear technologies is represented by their possible double usage: in the civilian and military fields, i.e. to supply power or weapons. This aspect heavily conditioned the USA’s energy policy, which, since the second half of 1970, excluded the fuel reprocessing and the development of fast reactors. Also the public opinion is sensitive to this argument. It is worth noticing that even the once-through cycle option is not completely safe from the proliferation point of view; in fact the plutonium is not destroyed, and in the future it could be retrieved from the disposal. Only recently the American Government is reconsidering the chance of reprocessing the fuel: the US Department

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of Energy launched in February 2006 the Global Nuclear Energy Partnership (GNEP, 2007). It is an international agreement to develop new technologies and fuel management aimed to reduce waste, produce more energy and minimize proliferation concerns. The key factors are (i) to develop a reprocessing technology such that the recycled plutonium will be still usable for nuclear fuel but not for nuclear weapons; (ii) to develop a nuclear fuel service program such that the partner nations will provide fuel to the developing countries in exchange for their commitment to forgo enrichment and reprocessing activities, which are the two possible routes for a national proliferation plan. Then, the idea of denaturing plutonium has regained interest. For instance, the plutonium that contains a suitable amount of 238 Pu becomes denatured, as the generated heat is enough to melt or to detonate the explosive included in a nuclear device. Following some studies (Kessler, 2007; Broeders and Kessler, 2007), the required fraction of 238Pu to make plutonium non-proliferant is likely to be a bit higher than 6% and it is relatively simple to be obtained, making use of re-enriched uranium and/or fuels containing a very little fraction of neptunium. It is worth reminding, however, that there is not a general consensus on the characteristics required by plutonium to be considered denatured (De Volpi, 1979; Lovins, 1980; Sagara et al., 2005). On the wait for future closed cycle systems as envisaged in Generation IV and GNEP initiatives, one most compatible solution with the sustainability, in particular the non-proliferation requirements, appears starting the reduction/elimination of the accumulated civilian and military Pu stocks, by introducing, in the mid-term, the utilization of inert matrix fuels (IMFs) in present commercial LWRs. 6. Uranium-free fuels In IMF plutonium is embedded in an U-free matrix so as to burn it without breeding any new plutonium by neutron capture in 238U. Thus a more efficient consumption of plutonium is achieved compared to MOX fuel. While reprocessing of spent UO2 fuel is the necessary step to optimise the fuel utilization, recycling second and third generation plutonium as MOX in LWRs is not favourable at least from an economic standpoint. This is a one main reason why a once-through cycle U-free IMF has been conceived for the plutonium burning. Suggestions of different types of IMF were based on neutronic, safety, non-proliferation as well as economic evaluations (OECD, 1999; Degueldre and Yamashita, 2003; Porta, 2000). To fit LWR requirements, the fuel assemblies should be designed just to replace conventional MOX elements, with addition of burnable absorbers like IFBA or erbium oxide to reduce power peaking factors. Loading up to 12 IMF assemblies in a large PWR core could be done by following fuel management schemes that are quite compatible with current operational requirements such as cycle length, soluble-boron concentration, etc. The addition of thoria has been considered because it is beneficial to cope with reduced reactivity feedbacks (low Doppler coefficient) in pure IMF (Lombardi et al., 1999b; Lombardi et al., 2001). The relative plutonium consumption rates for IMF assemblies, which were estimated to be up to 2.5 times higher than for MOX assemblies, represent a considerable incentive for PWR cores partially loaded with IMF (Kasemeyer et al., 1998). The other positive side of IMF utilization is that less Pu-fuel assemblies, with longer decay heat production, would be stored in the spent fuel reactor pool (Chawla and Konings, 2001). Development programmes focusing on material technology with the emphasis on fabrication, characterisation and irradiation of different IMF concepts have been started during the last decade in Europe, Japan and the USA. One of the earliest concept to be suggested among innovative fuel type IMF consists of plutonium

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dissolved in stabilised zirconia (ZrO2), a highly radiation-resistant material with good potential of achieving high burnup (Hellwig et al., 2006). The cubic structure of the zirconia matrix can be stabilized either by yttria (YSZ) or by calcia (CSZ) (Hellwig et al., 2006). The extreme insolubility of the matrix makes the fuel safe in case of cladding failure during irradiation (no washout of fissile material), and with regard to direct waste disposal. The YSZ and CSZ matrix is not redox sensitive, as it has to be assumed for uranium-based fuel in the safety assessment of geological underground storage (Hellwig et al., 2006). The fresh as well the spent zirconia based IMF is practically insoluble in aqueous and acidic solutions; this is an effective barrier against misuse of the fissile material. The major drawback of the homogeneous, solid solution, stabilized zirconia based fuel is the comparable low thermal conductivity (Hellwig et al., 2006). To cope with that, various countermeasures were discussed such as the application of annular pellets to reduce the fuel central temperature, and the adoption of a heterogeneous concept, like YSZ-based fuel particles embedded in a spinel matrix. Also the addition of thoria or urania not only for neutronic reasons but also to increase the thermal conductivity was discussed. Strictly speaking, the fuel is then no more a pure IMF though it will also be considered in this context. Two irradiation experiments related to the homogeneous concept were performed in the OECD-Halden Reactor in Norway in the years 2000–2005. The CSZ-based IMF, with and without thoria addition, fabricated according to the standard mixed drypowders route, was tested in IFA-652 (see Fig. 3), a joint Halden Programme experiment conducted in collaboration with ENEA, Italy (Hellwig et al., 2006; Streit et al., 2006). The YSZ-based IMF fabricated on two different routes was tested in IFA-651, a joint Halden Programme experiment conducted in collaboration with PSI (Switzerland) and Korea Atomic Energy Research Institute (KAERI) (Hellwig et al., 2006). 6.1. IFA-651 and 652 IMF experiments The main objective of both IMF experiments IFA-651 and 652 was to assess the basic behaviour of YSZ and CSZ-based IMF under irradiation conditions similar to those in current LWRs. Of particular interest was the measurement/evaluation of the fundamental variables like thermal conductivity and its degradation with burnup, fission gas release, fuel densification and swelling (Hellwig et al., 2006; Streit et al., 2006). For the inert matrix/MOX fuel test IFA-651 the secondary aim was to compare the performance of IMF with that of MOX. MOX and IMF pellets have mainly been fabricated in the Pu-laboratories of Paul Scherrer Institute. The primary topics of the inert matrix/thoria IFA-652 experiment were the same as for IFA-651. The secondary aim was to compare the performance of the two types of IMF with (Th,U)O2 fuel. All fuel pellets have been fabricated at IFE-Kjeller in Norway, by following the process and specifications set up by ENEA on the basis of cold pre-fabrication tests carried out in Italy (Vettraino et al., 1999). For practical reasons, high enriched uranium (HEU) was used as fissile material in this case, instead of Pu. Both IFA-651 and 652 contained a six rod cluster assembled and instrumented at IFE-Kjeller labs. Three rods with MOX, and three rods with IMF in IFA-651. Two rods for each of the three fuel types in IFA-652: CSZ-based IMF with HEU (instead of Pu), CSZ-based IMF with HEU and 39.2 wt% ThO2, and (Th,HEU)O2. In Fig. 4, these three fuel types are referred to by the acronyms IM, IMT and thoria, respectively. All fuel rods were equipped with fuel thermocouples and pressure transducers to measure the evolution of fuel temperature and pin inner pressure. Additionally three rods were equipped with fuel stack elongation sensors. The irradiation test for both rigs IFA-651 and 652 started successfully in June 2000 in the Halden reactor and the average assembly burnup reached at the end of irradiation in 2005 was in

Fig. 3. Schematic view of IFA-652 test-rig.

the range 340–480 kWd/cm3 (about 34–48 MWd/kgoxeq3), quite close and over the design target for both of them. After the end of the first loading (May 2003) one IMF rod (217 kWd/cm3 burnup) from IFA-651 was discharged for post-irradiation examination (PIE) as well as one (Th,HEU)O2 rod (190 kWd/cm3 burnup) from IFA-652 due to a clad welding defect. The irradiation history for IFA-652 is depicted in Fig. 4 where are reported as a function of the irradiation time (days and/or irradiation cycle) the assembly total power, the assembly average linear power (LHR) and the burnup at rod and assembly level. On the basis of experimental data from (Th,HEU)O2 rods of this irradiation test, the fuel performance code INFRA, which was developed at KAERI (Lee et al., 2001), has been first validated and therefore adopted to analyse the behaviour of the ThO2–UO2 fuel in a typical once-through cycle PWR. These analyses (Lee et al., 2004) show that: (i) such a fuel can probably be irradiated up to a burnup of 70–100 MWd/kgHM by optimizing the fuel rod design (e.g., by using annular pellets and a zirconium–niobium alloy with a higher corrosion resistance than standard Zircaloy-4);

3 Being the heavy metal content of an IMF much lower than the one of a standard fuel, it is usual to measure its burnup by the kWd/cm3. Another possibility is to report the value, measured by the MWd/kgoxeq, of a UO2 fuel with the same released energy per volume; the conversion is straightforward: 1 MWd/kgoxeq z 10 kWd/cm3.

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Fig. 4. Irradiation history of IFA-652.

(ii) its overall irradiation performance would be somewhat better than for UO2 fuel. The two experiments in Halden have shown a good irradiation response so as to put an important knowledge basis in a perspective of the U-free IMF viability in the current commercial LWRs and next generation reactors. Additions like yttria, calcia, erbia, and thoria, decided upon neutronic requirements and dissolved in the matrix, do not alter the irradiation behaviour significantly. A special effort, however, is still required on fabrication process and quality control side. Both fuels used for the irradiation experiments, YSZ and CSZ based, suffered from low density and/or unexpected density changes under irradiation. It is expected that the as-fabricated density can be strongly improved by using purified starting material and higher sintering temperatures. Finally, it is worth reminding that stabilised zirconia has been considered also for minor actinides transmutation (OECD/NEA, 2005), in fact this is one of the selected inert matrices in americium target fabrication either by using the classical powder metallurgy process or more advanced routes like the INRAM wet infiltration technique. The recent political developments (Generation IV and GNEP programmes) have in a way decreased chances for IMF deployment in an LWR environment as the call for fast and efficient plutonium destruction has slowed down, as a consequence of renewed interest in closed fuel cycle strategy. Nevertheless, IMF application/exploitation appears still of interest for next 20–30 years period, on the wait for deployment of future closed cycle systems, especially to cope with proliferation issues arising from the ever increasing weapons and civilian plutonium stocks. 7. Conclusions Nuclear is today a sustainable energy source under the criteria of economics, safety, availability of resources and impact on the

environment. The social acceptability is not granted, and the aspects linked to the environmental impact, as the recycle of wastes, will need to be more and more improved. But there is a further requisite for sustainability, peculiar to nuclear power, consisting in the resistance to nuclear armaments proliferation. It is an important aspect, which dictated the nuclear policy of the USA and influenced the one of other countries. The solution can be found through a combination of agreements among nations, modifications to fuel composition and improvements to reprocessing technology. The new nuclear – excluding fusion – is often identified by the public opinion with novel reactors and cycles, above all the reactor proposed by professor Rubbia (ADS) and thorium bearing fuels, which should represent a radical turn. From one side we are witness of a visionary attitude (dreaming machine and reprocessing), which proposes an accelerator driven subcritical fast reactor with thorium–uranium closed cycle þ new reprocessing technology þ totally remote fuel fabrication. This can solve a real problem, i.e. the proliferation risk, and the perceived ones, i.e. waste toxicity, plant safety, and optimal resources exploitation. From the other side we can find a radical rational attitude, claiming that the recycle is not justified: advantages are uncertain and far in the future, disadvantages are real and in the present. Our opinion is that the best way to follow is a co-existence of technologies: thermal and fast reactors; uranium, thorium and inert matrix bearing fuels; closed cycle in breeder reactors, semi-closed cycle with fissile topping, once-through cycle with deep fissile (and/or minor actinides) burning and fuel fit for the direct geological disposal. In particular, we focused on the latter option, as it can be implemented in present water reactors with relative small efforts and in a relatively quick time. The use of inert matrix and thoria fuels appears as an excellent short-term option to eliminate the proliferation risk associated to plutonium, and to lower its long-term radiotoxic inventory. The

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demonstration of the feasibility of inert matrix and thoria fuels will alleviate the concerns about these two issues, until a more efficient and sustainable nuclear technology will be available.

References Anthonysamy, S., Ananthasivan, K., Chandramouli, V., Kaliappan, I., Vasudeva Rao, P.R., 2000. Combustion synthesis of urania–thoria solid solutions. J. Nucl. Mater. 278, 346. Assmann H., Robin J.P., 1983. Guidebook on Quality Control of Mixed Oxides and Gadolinium Bearing Fuels for Light Water Reactors. International Atomic Energy Agency Report IAEA-TECDOC-584. Bakker, K., Cordfunke, E.H.P., Konings, R.J.M., Schram, R.P.C., 1997. Critical evaluation of the thermal properties of ThO2 and Th1yUyO2 and a survey of the literature data on Th1yPuyO2. J. Nucl. Mater. 250, 112. Balakrishna, P., Varma, B.P., Krishnan, T.S., Mohan, T.R.R., Ramakrishnan, P., 1998. Thorium oxide: calcination, compaction and sintering. J. Nucl. Mater. 160, 88. Basak U., Majumdar, S., Kamath, H.S., 2004. Development of UO2/MOX Fuels of Modified Microstructure for Improved Performance. International Atomic Energy Agency Report on Advanced Fuel Pellet Materials and Designs for Water Cooled Reactors IAEA-TECDOC-1416. Belle, J., Berman, R.M., 1984. Thorium Dioxide: Properties and Nuclear Applications. U.S. Department of Energy Report DOE/NE-060. Boussier, H., Delpech, M., Groullier, J.P., Renard, A., Volkaert, G., 2000. Evaluation of Possible Partitioning and Transmutation Strategies and of Means for Implementing Them. European Commission Report EUR 19128 EN, Luxembourg. Broeders, C.H.M., Kessler, G., 2007. Fuel cycle options for the production and utilization of denatured plutonium. Nucl. Sci. Eng. 156, 1. Cerrai, E., Lombardi, C., 1992. Burning weapon-grade Pu in ad hoc designed reactors. Nucl. Sci. Eng. 121, 1. Chawla, R., Konings, R.J.M., 2001. Categorisation and priorities for future research on inert matrix fuels: an extended synthesis of the panel discussions. Prog. Nucl. Energy 38, 455. Degueldre, C., Yamashita, T., 2003. Inert matrix fuel strategies in the nuclear fuel cycle: the status of the initiative efforts at the 8th inert matrix fuel workshop. J. Nucl. Mater. 319, 1. Demkowicz, P.A., Jerden Jr., J.L., Cunnane, J.C., Shibuya, N., Baney, R., Tulenko, J., 2004. Aqueous dissolution of urania–thoria nuclear fuel. Nucl. Technol. 147, 157. De Volpi, A., 1979. Proliferation, Plutonium and Policy. Pergamon Press, New York, USA. DOE, 1980. Nuclear Proliferation and Civilian Nuclear Power. DOE/NE-0001/1 through 9. Washington, DC DOE, 2002a. Yucca Mountain Site Suitability Evaluation. DOE/RW-0549, Washington, DC DOE, 2002b. A Technology Roadmap for Generation IV Nuclear Energy Systems. U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum Report GIF-002-00. DOE, 2003. Optimization of Heterogeneous Schemes for the Utilization of Thorium in PWRs to Enhance Proliferation Resistance and Reduce Waste. DOE/NE-0125, Washington, DC. Dziadosz, D., Ake, T.N., Saglam, M., Joe, J., Sapyta, J.J., 2004. Weapons-grade plutonium–thorium PWR assembly design and core safety analysis. Nucl. Technol. 147, 69. Fink, J.K., 2000. Thermophysical properties of uranium dioxide. J. Nucl. Mater. 279, 1. Fourest, B., Vincent, T., Lagarde, B., Hubert, S., Baudoin, P., 2000. Long-term behavior of a thorium-based fuel. J. Nucl. Mater. 282, 180. Galperin, A., Shwageraus, E., Todosow, M., 2002. Assessment of homogeneous thorium/uranium fuel for pressurized water reactors. Nucl. Technol. 138, 111. GNEP, 2007. http://www.gnep.energy.gov/pdfs/gnepSOP_091607.pdf. Goldberg, I., Waldman, L.A., Giovengo, J.F., Campbell, W.R., 1979. Fission Gas Release and Grain Growth in ThO2–UO2 Fuel Irradiated At High Temperature. Westinghouse Addendum WAPD-TM-1350. Gruppelaar, H., Kloosterman, J.L., Konings, R.J.M.,, 1998. Advanced Technologies for the Reduction of Nuclear Waste. Report ECN-R-98-008. ECN, Petten, Netherlands, ISBN 90-375-0003-X. Harada, Y., Baskin, Y., Handwerk, J.H., 1962. Calcination and sintering study of thoria. J. Am. Ceram. Soc. 45, 253. Hellwig, Ch., Streit, M., Blair, P., Tverberg, T., Klaassen, F.C., Schram, R.P.C., Vettraino, F., Yamashita, T., 2006. Inert matrix fuel behaviour in test irradiations. J. Nucl. Mater. 352, 291. Herring, J.S., MacDonald, P.E., Weaver, K.D., Kullberg, C., 2001. Low cost, proliferation resistant, uranium–thorium dioxide fuels for light water reactors. Nucl. Eng. Des. 203, 65. Herring, J.S., MacDonald, P.E., Weaver, K.D., 2004. Thorium-based transmuter fuels for light water reactors. Nucl. Technol. 147, 84. IAEA, 1970. Thorium Fuel Cycle. STI/PUB/21/39. In: International Atomic Energy Agency Bibliographical Series No. 39. IAEA, Vienna, Austria. IAEA, 1980. International Nuclear Fuel Cycle Evaluation. ISP534 1 through 9. IAEA, Vienna, Austria. IAEA-TECDOC-840, 1995. Unconventional Options for Plutonium Disposition. IAEA, Vienna, Austria. IAEA-TECDOC-916, 1996. Advanced Fuels with Reduced Actinide Generation. IAEA, Vienna, Austria.

IAEA-TECDOC-1155, 2000. Thorium Based Fuel Options for the Generation of Electricity: Developments in the 1990s. IAEA, Vienna, Austria. IAEA-TECDOC-1319, 2002. Thorium Fuel Utilization: Options and Trends. IAEA, Vienna, Austria. IAEA-TECDOC-1450, 2005. Thorium Fuel Cycle – Potential Benefits and Challenges. IAEA, Vienna, Austria. Iwamura, T., Uchikawa, S., Okubo, T., Kugo, T., Akie, H., Nakano, Y., Nakatsuka, T., 2006. Concept of innovative water reactor for flexible fuel cycle (FLWR). Nucl. Eng. Design 236, 1599. Joo, H.-K., Noh, J.-M., Yoo, J.-W., Cho, J.-Y., Park, S.-Y., Chang, M.-H., 2004. Alternative applications of homogeneous thoria-urania fuel in light water reactors to enhance the economics of the thorium fuel cycle. Nucl. Technol. 147, 37. J. Nucl. Mater. 274 (1–2), August 1999. J. Nucl. Mater. 319, June 2003 Complete. Kasemeyer, U., Chawla, R., Grimm, P., Paratte, J.M., 1998. Comparison of pressurized water reactor core characteristics for 100% plutonium-containing loadings. Nucl. Technol. 122, 52. Kessler, G., 2007. Plutonium denaturing by 238Pu. Nucl. Sci. Eng. 155, 53. Kim, H., Park, K., Kim, B.G., Choo, Y.S., Kim, K.S., Song, K.W., Hong, K.P., Kang, Y.H., Ho, K., 2004. Xenon diffusivity in thoria–urania fuel. Nucl. Technol. 147, 149. Kloosterman, J.L., 1998. Mutirecycling of Plutonium in Advanced PWRs. Netherlands Energy Research Foundation. ECN-R-98-007. Lahoda, E.J., 2004. Costs for manufacturing thorium–uranium dioxide fuels for light water reactors. Nucl. Technol. 147, 102. Lee, C.B., Bang, J.G., Kim, D.H., Kim, Y.M., Yang, Y.S., Jung, Y.H., 2001. Development of UO2 fuel rod performance analysis code INFRA. In: Korean Nuclear Society Topical Meeting, Suwon, October. Lee, C.B., Yang, Y.S., Kim, Y.M., Kim, D.H., Jung, Y.H., 2004. Irradiation behavior of thoria–urania fuel in a PWR. Nucl. Technol. 147, 140. Loewen, E.P., MacDonald, P.E., Hohorst, J.K., Sharp, J.P., Siefken, L.J., 2000. Development of Material Property Correlations for the Analysis of Mixed Thoria–Urania Fuel Performance in Light Water Reactors and Augmentation of the FRAPCON-3 Computer Code. Idaho National Engineering and Environmental Laboratory Report INEEL/EXT-2000-00275. Loewen, E.P., Wilson, R.D., Hohorst, J.K., Kumar, A.S., 2001. Preliminary FRAPCON-3Th steady-state fuel analysis of ThO2 and UO2 fuel mixtures. Nucl. Technol. 136, 267. Loewen, E.P., Weaver, K.D., Hohorst, J.K., 2002. Reactivity, isotopic, and thermal steady-state analysis of homogeneous thoria–urania fuel. Nucl. Technol. 137, 97. Lombardi, C., Padovani, E., Ricotti, M.E., Vettraino, F., 1999a. Plutonia–thoria fuel cycle as starting solution for a wider thorium use. In: Global’99 International Conference on Future Nuclear Systems, Jackson Hole, Wyoming, USA, August 29–September 3. Lombardi, C., Mazzola, A., Padovani, E., Ricotti, M.E., 1999b. Neutronic analysis of Ufree inert matrix and thoria fuels for plutonium disposition in pressurised water reactors. J. Nucl. Mater. 274, 181. Lombardi, C., Luzzi, L., Padovani, E., Vettraino, F., 2001. Inert matrix and thoria fuels for plutonium elimination. Prog. Nucl. Energy 38, 395. Long, Y., Yuan, Y., Kazimi, M.S., Ballinger, R.G., Pilat, E.E., 2002. A fission gas release model for high burnup LWR ThO2–UO2 fuel. Nucl. Technol. 138, 260. Long, Y., Siefken, L.J., Hejzlar, P., Loewen, E.P., Hohorst, J.K., MacDonald, P.E., Kazimi, M.S., 2004. The behavior of ThO2-based fuel rods during normal operation and transient events in LWRs. Nucl. Technol. 147, 120. Lovelock, J., 2006. The Revenge of Gaia. Allen Lane, UK. Lovins, A., 1980. Nuclear weapons and power-reactor plutonium. Nature 283 (5750), 817. Lucuta, P.G., Matzke, H.J., Verrall, R.A., 1994. Modelling of UO2-based SIMFUEL thermal conductivity – the effect of the burnup. J. Nucl. Mater. 217, 279. Lung M., 1997. A present review of the thorium nuclear fuel cycles. European Commission Report EUR 17771 EN, Luxembourg. Lung, M., Gremm, O., 1998. Perspectives on thorium fuel cycle. Nucl. Eng. Design 180, 133. MacDonald, P.E., 2000. Advanced Proliferation Resistant, Lower Cost, Uranium– Thorium dioxide Fuels for Light Water Reactors. Idaho National Engineering and Environmental Laboratory Report INEEL0EXT-2000-01217. MacDonald, P.E., 2001. Advanced Proliferation Resistant, Lower Cost, Uranium– Thorium dioxide Fuels for Light Water Reactors. Idaho National Engineering and Environmental Laboratory Report INEEL0EXT-01-00804. MacDonald, P.E., Lee, C.B., 2004. Use of thoria–urania fuels in PWRs: a general review of a NERI project to assess feasible core designs, economics, fabrication methodsPile Thermal/Mechanical Behavior, and Waste Form Characteristics. Nucl. Technol. 147, 1. Mah, B., 1983. Nuclear Reactor Materials and Applications. Van Nostrand Reinhold Company Inc., New York. NSESD, 1992. http://www.environment.gov.au/esd/national/nsesd/overview/index. html. OECD, 2002. Accelerator-Driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles. Nuclear Development Report No. 3109. OECD, Paris, France. OECD/NEA, 2005. Fuels and Materials for Transmutation. Report No. 5419. OECD/ NEA, Paris, France. OECD/NEA, IAEA, 2005. Uranium 2005: Resources, Production and Demand. OECD 662006031P1. OECD/NEA & IAEA. OECD Proceedings, 1999. In: Advanced Reactors with Innovative Fuels, Villigen, Switzerland, 21–23 Oct. 1998. ONU-IPCC, 2007. http://www.ipcc.ch/ipccreports/ar4-syr.htm. Pope, J.M., Radford, K.C., 1974. Physical properties of some thoria powders and their influence on sinterability. J. Nucl. Mater. 52, 241.

C. Lombardi et al. / Progress in Nuclear Energy 50 (2008) 944–953 Porta J., 2000. In: Fifth IMF Workshop, CRR-SERSI/ECP-99/30260, CEA DRN DER, Cadarache, France. Prog. Nucl. Energy 38 (3–4), 2001. Ronen, Y., 1990. High Converting Water Reactors. CRC Press, Boca Raton, Florida. Rubbia, et al., 1995a. Conceptual design of a fast neutron operated high power energy amplifier. CERN/AT/95-44 (ET). Rubbia, et al., 1995b. A realistic plutonium elimination scheme with fast energy amplifiers and thorium-plutonium fuel. CERN/AT/95-53 (ET). Sagara, H., Saito, M., Peryoga, Y., Ezoubtchenko, A., Takivayev, A., 2005. Denaturing of plutonium by transmutation of minor-actinides for enhancement of proliferation resistance. J. Nucl. Sci. Technol. 42, 161. Saglam, M., Sapyta, J.J., Spetz, S.W., Hassler, L.A., 2004. Core designs and economic analyses of homogeneous thoria–urania fuel in light water reactors. Nucl. Technol. 147, 8. Scott, D.S., 2004. Template for sustainability. Int. J. Hydrogen Energy 29, 17. Shwageraus, E., Zhao, X., Driscoll, M.J., Hejzlar, P., Kazimi, M.S., Herring, J.S., 2004a. Microheterogeneous thoria–urania fuels for pressurized water reactors. Nucl. Technol. 147, 20. Shwageraus, E., Hejzlar, P., Kazimi, M.S., 2004b. Use of thorium for transmutation of plutonium and minor actinides in PWRs. Nucl. Technol. 147, 53. Srirama Murti, P., Mathews, C.K., 1991. Thermal diffusivity and thermal conductivity studies on thorium–lanthanum mixed oxide solid solution. J. Phys. D 24, 2202. Streit, M., Tverberg, T., Wiesenack, W., Vettraino, F., 2006. Inert matrix and thoria fuel irradiation at an international research reactor. J. Nucl. Mater. 352, 263. Sunder, S., Miller, N.H., 2000. XPS and XRD studies of (Th,U)O2 fuel corrosion in water. J. Nucl. Mater. 279, 118.

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Suzuki, M., Saitou, H., Iwamura, T., 2004. Analysis of MOX fuel behavior in reducedmoderation water reactor by fuel performance code FEMAXI-RM. Nucl. Eng. Design 227, 19. ¨ nak, T., 2000. What is the potential use of thorium in the future energy production U technology? Prog. Nucl. Energy 37, 137. Vettraino, F., Magnani, G., La Torretta, T., Marmo, E., Coelli, S., Luzzi, L., Ossi, P., Zappa, G., 1999. Preliminary fabrication and characterisation of inert matrix and thoria fuels for plutonium disposition in light water reactors. J. Nucl. Mater. 274, 23. Wackernagel, M., Rees, W.E., 1996. Our Ecological Footprint: Reducing Human Impact on the Earth. New Society Publishers, Gabriola Island, BC. Weaver, K.D., Herring, J.S., 2003. Performance of thorium-based mixed-oxide fuels for the consumption of plutonium in current and advanced reactors. Nucl. Technol. 143, 22. Wilson, P.D., 2002. Development Requirements in Reprocessing Thorium Fuels. International Atomic Energy Agency Report IAEA-TECDOC-1319. Wilson, R., 2000. The changing need for a breeder reactor. Nucl. Energy 39 (2), 99. World Bank, 2002. World Development Report 2003: Sustainable Development in a Dynamic World: Transforming Institutions, Growth, and Quality of Life. Oxford University Press, New York, ISBN 0821351508. World Commission on Environment and Development (WCED), 1987. Our Common Future. Oxford University Press, New York, ISBN 0-19-282080-X. Yang, J.H., Kang, K.W., Song, K.W., Lee, C.B., Jung, Y.H., 2004. Fabrication and thermal conductivity of (Th,U)O2 pellets. Nucl. Technol. 147, 113. Zhao, X., Driscoll, M.J., Kazimi, M.S., 1999. Rationale for reconsidering the thorium cycle in light water reactors. Trans. Am. Nucl. Soc. 80, 43.