Energy Conversion and Management 51 (2010) 1774–1780

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Energy Conversion and Management

journal homepage: www.elsevier .com/locate /enconman

Evolution of nuclear fission reactors: Third generation and beyond

J.G. Marques *

Instituto Tecnológico e Nuclear, Estrada Nacional 10, 2686-953 Sacavém, Portugal

a r t i c l e i n f o

Article history:Available online 2 March 2010

Keywords:FissionNuclear energyNuclear reactorsAdvanced reactorsUranium resourcesRadioactive waste

0196-8904/$ - see front matter � 2010 Elsevier Ltd. Adoi:10.1016/j.enconman.2009.12.043

* Tel.: +351 219 946 115; fax: +351 219 941 039.E-mail address: jmarques@itn.pt

a b s t r a c t

Nuclear energy is attracting new interest around the world as countries look for low-carbon alternativesto fossil fuels to increase the diversity of their sources of energy and improve security of supply. Nuclearfission reactors provided approximately one sixth of the world’s electricity needs in recent years. The vastmajority of these reactors were built in the seventies and eighties. They are thus considered second gen-eration systems, as they are based on experience gained with the first generation or prototypes built inthe fifties and early sixties. Third generation reactors, developed in the nineties, are already a reality andwill dominate the market in the coming decades. A significant research effort is underway on systems ofthe fourth generation. Better economics, improved use of natural resources, less production of radioactivewaste, competitive production of hydrogen, and increased resistance to proliferation are within reachwith these new systems. A review will be done on the most important features of third and fourth gen-eration systems, together with a brief overview of the R&D challenges to be met.

� 2010 Elsevier Ltd. All rights reserved.

1. Introduction

Nuclear fission reactors provided approximately one sixth of theworld’s electricity needs in recent years, which translates intoabout 6% of the world’s primary energy needs [1]. Over 400 powernuclear fission reactors, with a total power of 270 GWe, are cur-rently in operation worldwide [2]. The vast majority of these reac-tors were built in the seventies and eighties. They are thusconsidered second generation systems, as they are based on expe-rience gained with the first generation or prototypes built in thefifties and early sixties. Nuclear fission reactors have a good trackrecord, reflected in over 12,000 reactor�years of accumulated expe-rience [3].

The safety of nuclear fission reactors has always been an impor-tant issue. The first fission reactor built by Fermi in 1942 [4] al-ready incorporated practical implementations of the basicconcepts of redundancy and diversity [5]. Defense-in-depth is asafety strategy in use by the nuclear industry since the fifties. It re-lies on having multiple, redundant, and independent layers ofsafety systems for the single, critical point: the reactor core. Itsapplication helps to ensure that the three basic safety functions –controlling the power and reactivity, cooling the fuel and confiningthe radioactive material – are preserved.

The progress of nuclear energy was strongly affected by theaccidents of Three Mile Island (USA) in 1979 and Chernobyl (Uk-raine) in 1986. Even if a core melt occurred in both cases, only

ll rights reserved.

the Chernobyl accident led to a significant collective dose tothe population [6] as most of the fission products inventorywas contained by the Three Mile Island design, in a clear dem-onstration of the effectiveness of the defense-in-depth principle.Mostly as a result of the Chernobyl accident, the public shows anextreme sensitivity to nuclear incidents, regardless of their realimpact.

New designs have a strong emphasis on improved safety andmore stringent safety objectives and requirements [5]. Thus, thirdgeneration designs achieve a significant reduction of radioactivereleases due to all conceivable accidents, including core melt acci-dents. Other issues, like a better use of uranium resources and amore rational handling of radioactive waste will only be addressedin the fourth generation designs, which will be a reality in the com-ing decades.

2. Generation III systems

Third generation reactors, developed in the nineties, are alreadya reality and will dominate the market in the next decades. Theirdesigns address the following issues [5,7]:

� Standardized designs, able to expedite licensing, reduce capitalcost and construction time.

� A simpler and more rugged design, making the reactors easier tooperate and less vulnerable to operational upsets.

� Greater use of passive safety features that require no active con-trols and rely on natural phenomena for corrective actions.

� Reduced probabilities of core melt accidents.

J.G. Marques / Energy Conversion and Management 51 (2010) 1774–1780 1775

� New mitigation measures in case of core melt accidents, in orderto reduce significantly the impact of such accidents to the envi-ronment and the populations.

� Resistance to aircraft impact.� Longer operating life, of 60 years.� Longer time between refueling, resulting in a higher availability.� Higher burn-up to reduce fuel use and the amount of waste.

Table 1 shows the main Generation III designs constructed, un-der construction or awaiting licensing [7,8]. The dates of expecteddeployment, when available, are also shown.

The Advanced Boiling Water Reactor (ABWR) developed byGeneral Electric (USA) and Hitachi (Japan) was the first GenerationIII reactor to enter commercial operation, in 1996. The ABWR hasinternal recirculation pumps inside of the reactor pressure vesselreplacing the external pumps of older Boiling Water Reactor(BWR) designs. This new design simplifies the connections to thevessel and reduces significantly the worst possible leak belowthe core region. In the event of a core melt, a basaltic floor withpassive cooling features will terminate the flow of corium [9]. FourABWR units are currently in operation in Japan and another threeunits are under construction in Taiwan and Japan. Two units areslated to start operation in the US in 2016–2017 [10].

The AES-92 is an advanced Pressurized Water Reactor (PWR) ofRussian design, with a power of 1.0 GWe [11], also designated NPP-92 or V-392 [12]. The main departure of the AES-92 from previous‘‘Vodo-Vodyanoi Energetichesky Reactor” (VVER) designs is the useof a combination of active and passive safety systems for increasedreliability. The new design has a planned lifetime of 60 years. Acore catcher will be provided as a mitigation measure [11,13].Two units are currently being built in India [7].

The European Pressurized Reactor (EPR) is an evolution of theFrench ‘‘N4” and German ‘‘Konvoi” reactors. It has four indepen-dent cooling systems, an extra cooling and containment area inthe bottom to catch the molten core if a core meltdown should oc-cur [14] and a containment building that can withstand a directcrash by a large airplane. Although designed for UO2 fuel, theEPR core can use up to 50% Mixed OXide (MOX, a blend of oxidesof plutonium and uranium) fuel [15]. Its power is 1.6 GWe, makingit one of the largest plants available. Two EPR units are under con-struction, one in Finland and another in France. The construction oftwo additional units is planned to start in 2009 in China.

Westinghouse (USA) has a new series of advanced passive PWR,available in two models – the AP600 at 600 MWe and the AP1000at 1000 MWe [16]. The AP600 got its final design approval fromthe Nuclear Regulatory Commission (NRC) in the USA in 1998,but no units were built. The AP1000 is basically a scaled-up ver-sion of the AP600 and was certified by the NRC in 2005. The safetysystems apply passive protection and thus no emergency dieselgenerators are required in the case of a loss of electrical supply.In the event of a core melt the operator can flood the reactor cavityspace immediately surrounding the reactor vessel with water tosubmerge the reactor vessel; this cooling is sufficient to prevent

Table 1Generation III reactors under construction or undergoing a licensing procedure.

Reactor Developer(s) Po

ABWR General Electric, Hitachi 1.AES-92 Gidropress 1.EPR Areva 1.AP600, AP1000 Westinghouse 0.APR-1400 KHNP, Westinghouse 1.APWR Mitsubishi 1.ESBWR General Electric 1.ACR-1000 Atomic Energy of Canada 1.

a Expected.

molten core debris in the lower head from melting the steel vesselwall and spilling into the containment. The design is less expen-sive to build than a conventional PWR due to a significant reduc-tion in the number of pipes, wires, valves and associatedcomponents. The target construction time for the AP1000 is only36 months. The construction of the two AP1000 units in Chinastarted in 2009 [17].

The System 80+ reactor is a PWR designed by Combustion Engi-neering (USA) and by its successor owners Asea Brown Boveri(Switzerland) and Westinghouse. It was certified by the NRC in1997, but no units were built. System 80+ provided a basis forthe new APR1400 design in South Korea. The APR1400 is a1.4 GWe PWR, equipped with passive safety measures [18] and acorium retention system [19]. The first unit is expected to startcommercial exploitation in South Korea in 2013.

The Advanced Pressurized Water Reactor (APWR) was devel-oped by Mitsubishi (Japan). It features several design enhance-ments including a neutron reflector, improved efficiency andimproved safety systems. It uses a combination of passive and ac-tive safety systems [20]. The first APWR with a power of 1.5 GWe isexpected to start commercial operation in Japan in 2016. Subse-quent units are expected to have a power of 1.70 GWe and fullMOX core abilities.

General Electric began in 1992 a project to design and certify aBWR design which incorporated advanced, passive safety features.The Simplified Boiling Water Reactor was a natural circulationreactor rated at 0.67 GWe. This design ultimately evolved intothe 1.55 GWe Economic Simplified Boiling Water Reactor (ESBWR)design. The ESBWR has no recirculation pumps, thereby greatlyincreasing design integrity and reducing overall costs. The pas-sively safe characteristics are mainly based on isolation condens-ers, which are heat exchangers that take steam from the vesselor the containment, condense the steam, transfer the heat to awater pool, and introduce the water into the vessel again. All safetysystems operate without using pumps, thereby further increasingdesign safety reliability and reducing costs. The core is made short-er than conventional BWR plants to reduce the pressure drop overthe fuel and improve natural circulation [21].

Atomic Energy of Canada Limited has developed an improved‘‘CANada Deuterium Uranium” (CANDU) design. The AdvancedCANDU Reactor (ACR) uses slightly enriched uranium fuel(0.9%235U), instead of natural uranium as previous models, whichis expected to result in operational savings [22]. The ACR will havea small negative coolant void coefficient [23], in contrast with ear-lier CANDU designs, which feature a small positive void coefficient[24]. The coolant of the ACR will be light water instead of heavywater, which will be retained only as moderator. The new designwill simplify the complex system of cooling pipes running througha more compact core and will use new alloys for the piping, in or-der to guarantee a lifetime of 60 years. The ACR has a planned life-time capacity factor greater than 90% [25]. The first ACR-1000, witha power of 1.2 GWe is expected to be operational in Canada around2016.

wer (GWe) Type First deployment

37 BWR Japan, 199600 PWR India, 2009a

60 PWR Finland, 2012a

65, 1.12 PWR China, 2013a

30 PWR South Korea, 2013a

50–1.70 PWR Japan, 2016a

55 BWR Awaiting license20 PHWR Awaiting license

Fig. 2. Use of uranium resources at current consumption levels in three scenarios:(left) open fuel cycle and only identified uranium resources, (middle) open fuelcycle and all conventional and non-conventional uranium resources, (right) closedfuel cycle, with fast reactors and all conventional and non-conventional uraniumresources.

1776 J.G. Marques / Energy Conversion and Management 51 (2010) 1774–1780

3. Uranium resources and waste management

The vast majority of today’s fission reactors operate in an openfuel cycle, in which uranium is mined, eventually enriched in 235U,partially fissioned in the reactor and stored after that. In such a cy-cle, a generic 1.0 GWe Light Water Reactor (LWR) producing6.6 TWh per year requires about 150 tons of natural uranium,which is transformed into 27 tons of enriched uranium, out ofwhich approximately only 1 ton will be fissioned [26]. The averagedischarge burn-up rate has significantly increased in the last dec-ades, as shown in Fig. 1, which displays non-pondered averagesof the values of Watteau and co-authors [27] for PWR and BWRreactors in different regions of the world. The average dischargeburn-up rate will be further increased with new designs – e.g.,for the EPR it will be at least 55 GWd/tHM [15], which is nearlythe double of the value achieved in the early eighties. However,there is no clear economic gain when increasing the average dis-charge burn-up rate above 60 GWd/tHM [28] with current reac-tors, even if this would be technically feasible. These limitationsmake it obvious that an open fuel cycle is a waste of usefulresources.

The status of uranium resources is periodically published in theso called ‘‘Red Book” [2]. The sum of the identified resources is cur-rently 5.5 Mt, with an estimated recovery cost below 130 USD/kg.It is expected that 10.5 Mt may still be added to the above value,which would bring the total conventional uranium resources to16 Mt. Furthermore, the ‘‘Red Book” includes an estimate of theamount of uranium that could be recovered from phosphates(which have 50–200 ppm of U), also with proven technology, atabout 22 Mt. Therefore, the highest value that can be currently as-sumed for uranium resources worldwide is 38 Mt.

Fig. 2 illustrates the time scale of usage of the uranium re-sources in three scenarios. If the uranium consumption would re-main at the current level of 60–70 kt per year [2], with a fleetdominated by PWR and BWR reactors, there would be enoughore for 70–80 years if we take only the identified resources orabout 500 years if we consider all resources.

Two basic measures can be taken to increase the time span ofuranium resources:

� Closure of the nuclear fuel cycle, through reprocessing of spentfuel and introduction of recovered uranium and plutonium intonew fuel.

� Introduction of reactors using fast neutrons (so called fast reac-tors) which can effectively use 238U, the most abundant uraniumisotope.

Fig. 1. Average discharge burn-up rates of BWR and PWR fuels, from 1980 to 2000.

Both measures are well known, as they were developed in theearly days of nuclear energy, when the known uranium resourceswere significantly lower. An increase by a factor of 30 was usedin Fig. 2 to account for the closure of the cycle and introductionof fast reactors.

The closure of the fuel cycle is a reality in several countries,notably France, where more than 1600 Mt of spent fuel are repro-cessed every year [29]. Recycled uranium can be brought back intothe UO2 fuel fabrication chain. Separated plutonium can be used inMOX fuel for the current fleet of reactors based on fission by ther-mal neutrons. MOX fuel is routinely used in about 30 reactors inEurope [30] and its use is increasing significantly: 16 metric tonsin 2008 while it was 8 metric tons the year before [31]. A PWRusing 30% MOX fuel consumes an amount of Pu comparable tothe one that is produced in the other assemblies using enricheduranium [32]. MOX fuel can also be a solution to dispose of surplusweapons grade plutonium – the USA is currently constructing aMOX fabrication facility for this end [33]. On the negative side,reprocessing raises some concerns on proliferation [34].

Fast neutron reactors have the ability to use 238U, the dominantisotope in natural uranium. The first fast neutron reactor, Clemen-tine, was built in Los Alamos in 1946 [35]. Fast reactor designsrange from ‘‘breeders” that can produce more usable nuclear fuelthan they consume (i.e., with a high conversion ratio) to ‘‘burners”(i.e., producing less new fuel but consuming more transuranic ele-ments, with a low conversion ratio). About 20 fast reactors have al-ready been operating in several countries, with an accumulatedexperience of 300 reactor�years [36]. Although the availability ofuranium at reasonably low prices has made less urgent the intro-duction of fast reactors, a renewed interest emerged meanwhiledue to their ability to transmute the longest lived isotopes con-tained in nuclear waste. Transmutation is here used to includethe conversion of long-lived isotopes into other isotopes with ashorter half-life through neutron capture, as well as the fission ofthe actinides, which is frequently called ‘‘incineration”.

The management and disposal of radioactive waste arising fromthe nuclear power industry is a delicate problem. After more thanfive decades of nuclear energy, no country has yet succeeded indisposing of its high-level nuclear waste. The preferred technolog-ical solution is the use of repositories constructed in deep stablerock formations. This is, to some extent, a solution demonstratedby the Oklo natural reactors [37] whose fission products migratedvery little in a time scale of the order of 1 billion years [38], evenwithout engineered barriers. However, the public acceptance of

J.G. Marques / Energy Conversion and Management 51 (2010) 1774–1780 1777

this solution is far from guaranteed – while the creation of a repos-itory in Italy was recently abandoned due to strong public opposi-tion [39], it progresses smoothly in other European countries, likeFinland and Sweden.

A recent study [34], which advocates the use of an open fuel cy-cle for non-proliferation reasons, concludes that a repositoryequivalent to the Yucca Mountain (USA) would be necessaryworldwide every 3–4 years in a scenario with 1000 GWe of nuclearpower, roughly four times the currently installed value. Thusrepository space can be a serious bottleneck.

The relative radiotoxicity of unprocessed spent fuel only be-comes equivalent to the one of natural uranium after a decay per-iod of 0.2 million years [32]. This period decreases by an order ofmagnitude if uranium and plutonium are separated and only thefission products and minor actinides (MA) – neptunium, ameri-cium and curium – are stored. It further decreases to 300 years ifonly the fission products are stored.

The MA can be recycled in the existing fleet of LWR togetherwith plutonium. However, this is not the best solution, as it is nec-essary to compensate the ‘‘poison” effect of the MA and, since acti-vation of the MA predominates over their fission in a thermalneutron spectrum, a high curium content ends up being produced,which severely complicates handling of the fuel [40]. Fast reactorsare more efficient in this regard: for the same plutonium consump-tion the inventory of MA is three times smaller than on a LWR [41].

Accelerator Driven Sub-critical (ADS) systems provide the pos-sibility of using ‘‘dedicated” cores, where the fuel can be heavilyloaded with MA. ADS systems couple a sub-critical core with anaccelerator which provides extra neutrons obtained by spallationreactions using a high-energy proton beam. Although the roots ofADS can be traced back as far as the forties [42], this type ofarrangement became workable only in the early nineties [43–45]with the progress in accelerator technology. ADS offer the featureof sub-criticality (with a multiplication factor around 0.95–0.99)which overcomes the drawbacks of the low delayed neutron frac-tion (from the MA) and reduced Doppler Effect (due to the absenceof 238U) [46] which could be problematic in a fast reactor.

MYRRHA is an ADS under development in Belgium [47]. It aimsto serve as a basis for the European eXperimental demonstration ofTransmutation in ADS (XT-ADS) program and to provide protonsand neutrons for various research and development applications.It consists of a proton accelerator delivering a 350 MeV to 5 mAbeam to a liquid Pb–Bi spallation target that in turn couples to aPb–Bi cooled, sub-critical fast nuclear core [48]. It is expected thatMYRRHA will be in service in 2016–2018 as the first operationalADS.

The importance of partitioning and transmutation in wastemanagement is widely recognized. Several scenarios using ADSand/or fast reactors in complement to a fleet of LWR are possible[41,49]. A significant number of studies have been done recentlyor are in progress at both national and international level. The ‘‘Ad-vanced Fuel Cycle Initiative” of the Department of Energy of theUSA proposed the introduction of limited recycling using LWR by2025 (in which spent fuel is recycled a few times before being dis-posed), followed by a transitional recycling phase using fast and

Table 2Generation IV systems under development.

Reactor Spectrum

Sodium-cooled Fast Reactor (SFR) FastLead Alloy Cooled Reactor (LFR) FastGas-cooled Fast Reactor (GFR) FastVery High Temperature Reactor (VHTR) ThermalSupercritical Water Cooled Reactor (SCWR) Thermal and fasMolten Salt Reactor (MSR) Thermal

LWR by 2040 (in which spent fuel is recycled continuously andonly fission products are disposed) and ending in a sustained recy-cling infrastructure by 2100 [50]. Views and strategies of othercountries, such as France [51] and Japan [52] were presented in ameeting recently convened by the International Atomic EnergyAgency (IAEA). At a regional level, a recent study published by Nu-clear Energy Agency (NEA) shows the clear benefits of the adoptionof a partitioning and transmutation strategy in Europe, even whenthe countries have different nuclear policies [53].

The IAEA [54,55], NEA [40,56–58] and the European Commis-sion [59,60] are taking active roles promoting studies in comple-mentary aspects of partitioning and transmutation and in thetransition to sustained recycling. CERN, where ADS received a sig-nificant boost in the early nineties, supports the important n_TOFexperiment [61], where nuclear data of primordial importance istaken.

4. Generation IV systems

While third generation designs have a strong emphasis on im-proved safety and more stringent safety objectives and require-ments, fourth generation systems are expected to showsignificant progress in economy, safety, environmental perfor-mance, and proliferation resistance.

The Generation IV International Forum (GIF) was established in2001 with the aim of fostering the research and development nec-essary for these new nuclear energy systems. The GIF published atthe end of 2002 an extensive report [62], outlining six reactor con-cepts, briefly outlined in Table 2. The basic concepts of the GIF sys-tems are not entirely new. Although at different levels, all werepreviously studied and some prototype systems were even oper-ated. Three aspects of the GIF designs deserve to be highlightedas they represent significant departures from previous systems:

� Most designs consider closed fuel cycles.� Three of the six designs are fast reactors.� Four of the designs are high-temperature reactors and thus can

be used to produce hydrogen or synthetic fuels and have animpact on the transport sector.

The Very High-Temperature Reactor (VHTR) is a graphite-mod-erated, helium cooled reactor, planned for an open fuel cycle. Itsprimary design uses a 0.6 GWe thermal core with a helium outlettemperature up to 1000 �C. It is targeted for hydrogen productionusing, e.g., the iodine–sulfur cycle [63]. The estimated price ofhydrogen produced using a VHTR is 1.5–2 USD/kg [7].

Several high-temperature reactors are currently being activelydeveloped. The Pebble Bed Modular Reactor (PBMR) and the GasTurbine Modular Helium Reactor (GT-MHR) are representativeexamples of this class of reactors. These are not Generation IV de-signs but rather Generation III + designs. The PBMR is being devel-oped by the companies Eskom (South Africa), BNFL (UK) andExelon (USA). It features a pebble bed core and a closed cycle gasturbine. Its thermal power will be 400 MWt for an output of

Fuel cycle Temp. (�C)

Closed �550Closed 550–800Closed �850Open >900

t Open and closed 350–620Closed 700–800

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165 MWe, thus achieving an efficiency of 41%, operating at 900 �C[7]. The GT-MHR is being developed by General Atomics and byOKBM (Russian Federation). It features a prismatic core and closedcycle gas turbine. Its thermal power is expected to be 600 MWt foran output of 280 MWe; this design has thus 47% efficiency,achieved at an operating temperature of 1560 �F [7]. These reactorswill use fuel that is significantly different from the one used in cur-rent reactors: 1 mm-diameter spheres containing a kernel of fuelsurrounded by three CVD layers of carbon and one layer of siliconcarbide, in a design called tri-isotropic (TRISO)-layered particles[64], developed in Germany within a program started in the sixties[65].

The Gas-cooled Fast Reactor (GFR), Sodium-cooled Fast Reactor(SFR) and Lead-cooled Fast Reactor (LFR) are three GIF conceptsthat bridge the gap for fast reactors. The GIF considers that fastreactors can increase the transformation of mined reserves intouseful heat by two orders of magnitude when compared with ther-mal reactors on an open fuel cycle [62]. The GFR builds upon theGT-MHR design for a single combined turbine and compressorshaft. The SFR is probably the most mature of the GIF designs,building upon the experience acquired in several countries. TheSFR will operate at a temperature of 550 �C and could be deployedwith a power in the range of 0.15–0.50 GWe, or in the range of0.50–1.50 GWe. The LFR builds upon the experience obtained inlead–bismuth cooled fast reactors used for naval propulsion [66].The outlet temperature of the LFR is expected to be in the 550–850 �C range. The LFR could also be used to produce hydrogen ifits output temperature is close to the upper limit given above.The LFR could be deployed in a wide range of powers, from 0.05to 0.15 GWe with long refueling times (similar to a naval propul-sion reactor), to a large reactor at about 1.2 GWe.

The Molten Salt Reactor (MSR) does not use regular fuel assem-blies. Instead, the liquid molten fuel composed of uranium and plu-tonium fluorides dissolved in sodium or zirconium fluorides ispumped through the core. Research on this type of reactors wasstarted in the fifties in the USA, with the intention to power air-craft. An 8 MWth prototype was operated in Oak Ridge (USA) inthe late sixties [67]. The MSR will operate at a temperature of700–850 �C, thus enabling the production of hydrogen. The GIF isalso working on the design of a fast spectrum MSR that will havelarge negative temperature and void reactivity coefficients, a un-ique safety characteristic not found in solid-fuel fast reactors [68].

The Supercritical Water Cooled Reactor (SCWR) relies on the useof water that is above its critical point; physically steam and liquidare then indistinguishable. The goal is that more of the heat re-leased from fission is converted into electricity; the GIF estimatesan efficiency of 44% for the SCWR. Additionally, the devices han-dling the phase changes of water in conventional LWR are notneeded, which will improve reliability and safety, while reducingconstruction costs. There is a significant experience accumulatedwith supercritical fossil fuel fired boilers.

Fig. 3. Operating temperatures and expected displacement damage (in displace-ments per atom, dpa) of Generation IV designs compared with previousgenerations.

5. Generation IV R&D challenges

Several research and development (R&D) challenges lie ahead.Two of the most significant are the fuel compositions with theirimpact in the safety of the GIF reactors and the materials requiredfor those reactors [62,69].

The fuel compositions to be used in the GIF reactors are still anopen issue. A significant amount of work is underway on fuels, aswell as on separation techniques for reprocessing [64,70]. It shouldbe emphasized that plutonium will not be separated from otheractinides in all GIF systems that use a closed fuel cycle, in orderto make this nuclear material less attractive for a possible diver-sion [62].

The amount of MA in the fuel for fast reactors is limited due toits detrimental effect in the reactivity and kinetic coefficients. With4% MA on the fuel of a Sodium cooled fast reactor there is an in-crease of 10% in the coolant void coefficient, a decrease of slightlyless than 10% in the fraction of delayed neutrons and a decrease of30% in the Doppler coefficient. The above detrimental variationslimited the amount of MA in the fuel of the French fast reactor‘‘Superphénix” to 3% to keep acceptable core parameters [71].

The positive void effects in Sodium cooled fast reactors can bedecreased by at least an order of magnitude with new designsusing dense fuels (mono-nitride, carbide or metallic fuels), ele-vated fuel in-core fractions (where parasitic neutron capture is sig-nificantly depressed) and optimal core-blanket configurations [72].

The materials used in today’s commercial reactors are not suit-able for the GIF reactors. Fig. 3 illustrates the operating tempera-tures and expected displacement damage for the GIF reactors andfor older designs, using data from Guérin and co-authors [73]. Themain challenges to be faced are higher operating temperatures,higher radiation and more demanding coolants.

While the operating temperature of most current reactors isclose to 400 �C, a significant number of the new reactors will oper-ate at over 800 �C. High neutron fluences are expected for the fastreactors, with displacement damages up to 200 dpa (displace-ments per atom), while for previous generations this value is below75 dpa. In order to achieve a higher thermodynamic efficiency, thenew designs will use as coolants water in supercritical state, liquidmetals (sodium, bismuth or bismuth alloys), molten salts, and highpressure helium gas, which are more demanding coolants than or-dinary water. For most of the above environments, corrosion resis-tance critically depends on the ability to maintain suitableprotective surface layers, which can also be affected by radiation.Materials technology is thus the key to the success [74]. The mod-eling of materials behavior is a promising tool that can be used toovercome, at least partially, expensive trial-and-error experiments[75].

The candidate materials for GIF systems are ferritic–martensiticsteels, nickel based alloys, oxide dispersion strengthened ferritic–martensitic steels and ceramics [76]. Ceramic materials may be asolution for various components operating above 1000 �C.

Finally, one should note that the R&D needs for the implemen-tation of the GIF reactors are not decoupled from the needs tomaintain the current fleet of LWR and eventually extend its life-time. Europe has recently launched a ‘‘Sustainable Nuclear EnergyTechnology Platform” with over 70 stakeholders from industry, re-

J.G. Marques / Energy Conversion and Management 51 (2010) 1774–1780 1779

search and academia, technical safety organizations and non-gov-ernmental organizations. This platform has established a prelimin-ary research roadmap that integrates LWR of the second and thirdgenerations, fast reactors of the fourth generation, as well as newapplications of nuclear energy, in a time scale going as far as theyear 2040 [77].

6. Conclusions

Nuclear fission reactors are based on mature technology, with agood track record. The construction of new fission reactors isclearly increasing in several countries, after two decades of re-duced construction. Third generation reactors, developed in thenineties, are already a reality and will dominate the market inthe next two or three decades. Their designs have better safetyand efficiency than previous generations, thus positively address-ing the major concerns of the post-Chernobyl era.

A better use of natural resources and a decrease in the amountof generated waste will be done with nuclear systems of the fourthgeneration, currently under development. Fission has an affordableresource base of uranium ore good for millennia, provided that thefuel cycle is closed and fast neutron reactors are introduced. Thewaste generated by fission reactors can achieve a neutral radiotox-icity exchange with the earth’s crust by recycling and fissioning alltransuranic elements and sequestering fission products for aboutthree centuries.

References

[1] International Energy Agency. Key world energy statistics 2008. Paris: IEA;2008.

[2] Nuclear Energy Agency. Uranium 2007: resources, production and demand.Paris: NEA; 2008. ISBN 978-92-64-04766-2.

[3] World Nuclear Association. Safety of nuclear power reactors. Informationpaper no. 6. WNA; 2008. <http://www.world-nuclear.org/>.

[4] Glasstone S. Principles of nuclear reactor engineering. 1st ed. Princeton: VanNostrand; 1955. p. 788–91.

[5] Frisch W, Gros G. Improving the safety of future nuclear fission power plants.Fusion Eng Des 2001;56–57:83–93.

[6] Cooper JR, Randle K, Sokhi RS. Radioactive releases in the environment: impactand assessment. Chichester: Wiley; 2003. p. 149–62. ISBN 0-471-89924-0.

[7] Hore-Lacy I. Nuclear energy in the 21st century. London: World NuclearUniversity Press; 2006. p. 64–8. ISBN 0-12-373622-6.

[8] Energy Information Administration. New commercial reactor designs. EIA;2006. <http://www.eia.doe.gov/>.

[9] General Electric Hitachi. Advanced boiling water reactor fact sheet. GEA-14576E. GE Hitachi; 2008. <http://www.gepower.com/>.

[10] Nuclear Engineering International. EPC contract signed for South TexasABWRs; February 2009. <http://www.neimagazine.com/>.

[11] Ermakov Y, Rousselot O. EUR volume 3 AES 92 subset. European utility requirementsseminar 2007. Nice; 2007. <http://www.europeanutilityrequirements.org/>.

[12] International Atomic Energy Agency. Status of advanced light water reactordesigns 2004. Tecdoc-1391. Vienna: IAEA; 2004. p. 337–58. ISBN 92-0-104804-1.

[13] Song JH, Suh ND. An evolution of molten core cooling strategies. Nucl Eng Des2009;239:1338–44.

[14] Fisher M. The severe accident mitigation concept and the design measures forcore melt retention of the European pressurized reactor. Nucl Eng Des2004;230:169–80.

[15] Sengler G, Forêt F, Schlosser G, Lisdat R, Stelletta S. EPR core design. Nucl EngDes 1999;187:79–119.

[16] Paulson CK. AP1000: set to compete. Nuclear Engineering International;October 2002. <http://www.neimagazine.com/>.

[17] Nuclear Engineering International. First concrete at China’s haiyang;September 2009. <http://www.neimagazine.com/>.

[18] Chu IC, Song CH, Cho BH, Park JK. Development of passive flow controllingsafety injection tank for APR1400. Nucl Eng Des 2008;238:200–6.

[19] Kang KH, Park RJ, Kim SK, Suh KY, Cheung FB, Remped JL. Simulant meltexperiments on performance of the in-vessel core catcher. Nucl Eng Des2007;237:1803–13.

[20] Tujikura Y, Oshibe T, Kijima K, Tabuchi K. Development of passive safetysystems for next generation PWR in Japan. Nucl Eng Des 2000;201:61–70.

[21] Hinds D, Maslak C. Next generation nuclear energy: the ESBWR. Nucl News2006;49(1):35–40.

[22] Torgeson DF, Shalaby BA, Pang S. CANDU technology for generation III+ and IVreactors. Nucl Eng Des 2006;236:1565–72.

[23] Atomic Energy of Canada Ltd. ACR-1000 technical summary. AECL; 2007.<http://www.aecl.ca/>.

[24] Snell VG, Howieson JQ. Chernobyl – a Canadian perspective. AECL; 1999.<http://canteach.candu.org/>.

[25] Petrunik K. ACR-1000 ready to deliver: evolutionary technology experiencedteam and proven track record. In: Nuclear technology international 2008.London: Sovereign Publications; 2008.

[26] Lamarsh JR, Baratta AJ. Introduction to nuclear engineering. 3rd ed. PrenticeHall: Upper Saddle River; 2001. p. 187. ISBN 0-201-82498-1.

[27] Watteau M, Estève B, Güldner R, Hoffman R. Framatome ANP extended burnupexperience and views on LWR fuels. In: Proceedings of world nuclear associationannual symposium, London; 2001. < http://www.world-nuclear.org/>.

[28] Nuclear Energy Agency. Very high burn-ups in light water reactors. Paris: NEA;2006. ISBN 92-64-02303-8.

[29] Frois B. Nuclear energy in a global context. Nucl Phys A 2008;805:320c–7c.[30] Nuclear Decommissioning Authority. NDA plutonium options. NDA; 2008.

<http://www.nda.gov.uk/>.[31] European Commission. Euratom supply agency annual report 2008.

Luxembourg: Publications Office of the European Union; 2009. p. 29–30.ISBN 92-79-12561-4.

[32] Comissariat à L’Énergie Atomique. Les Combustibles Nucléaires. Paris: CEASaclay and Groupe Moniteur; 2008. p. 77–80. ISBN 2-281-11325-9.

[33] Nuclear Engineering International. US MOX plant construction begins; August2007. <http://www.neimagazine.com/>.

[34] Ansolabehere S, Deutch J, Driscoll M, Gray PE, Holdren JP, Joskow PL, et al. Thefuture of nuclear power: an interdisciplinary MIT study. Boston: MIT; 2003.ISBN 0-615-12420-8.

[35] Glasstone S. Principles of nuclear reactor engineering. 1st ed. Princeton: VanNostrand; 1955. p. 832.

[36] World Nuclear Association. Fast neutron reactors. Information paper no. 98.WNA; 2009. <http://www.world-nuclear.org/>.

[37] Neully M, Bussac J, Frèrejacques C, Nief G, Vendryes G, Yvon J. Sur L’existencedans un Passé Reculé d’une Reaction en Chaine Naturelle de Fissions, dans leGisement d’Uranium d’Oklo (Gabon). Comptes Rendus Acad Sci Paris1972;275:1847–9.

[38] Cowan GA. A natural fission reactor. Sci Am 1976;235:36–47.[39] Gunter LC. Nuclear waste: showdown at Scanzano. Bull Atom Sci

2004;60(2):12–3.[40] Nuclear Energy Agency. Physics and safety of transmutation systems: a status

report. Paris: NEA; 2006. ISBN 92-64-01082-3.[41] Salvatores M. Nuclear fuel cycle strategies including partitioning and

transmutation. Nucl Eng Des 2005;235:805–16.[42] European Technical Working Group on ADS. A European roadmap for

developing accelerator driven systems (ADS) for nuclear waste incineration.Rome: ENEA; 2001. ISBN 88-8286-008-6.

[43] Bowman CD, Arthur ED, Lisowski PW, et al. Nuclear energy generation andwaste transmutation using an accelerator-driven intense thermal neutronsource. Nucl Instr Meth Phys Res A 1992;320:336–67.

[44] Carminati F, Klapisch R, Revol JPC, Roche C, Rubio JA, Rubbia C. An energyamplifier for cleaner and inexhaustible nuclear energy production driven by aparticle beam accelerator. CERN-AT-93-47-ET. Geneva: CERN; 1993.

[45] Rubbia C, Buono S, González E, Kadi Y, Rubio JA. A realistic plutoniumelimination scheme with fast energy amplifiers and thorium–plutonium fuel.CERN-AT-95-53 ET. Geneva: CERN; 1995.

[46] Salvatores M. The physics of transmutation in critical or subcritical reactors.CR Phys 2002;3:999–1012.

[47] Aït Abderrahim H, Kupschus P, Malambu E, Benoit Ph, Van Tichelen K, Arien B,et al. MYRRHA: a multipurpose accelerator driven system for research &development. Nucl Instr Meth Phys Res A 2001;463:487–94.

[48] Maes D. Mechanical design of the small-scale experimental ADS: MYRRHA.Energy Convers Manage 2006;47:2710–23.

[49] Gudowski W. Nuclear waste management: status, prospects and hopes. NuclPhys A 2005;752:623c–32c.

[50] US Department of Energy. Advanced fuel cycle initiative: objectives, approach,and technology summary – report to congress. Washington: DOE; 2005.<http://afci.sandia.gov/>.

[51] Proust E, Debes M, Sire J-M. The benefit of reprocessing/recycling forsustainable nuclear energy: the French view. In: Proceedings of technicalmeeting on fissile material management strategies for sustainable nuclearenergy. STI/PUB/1288. Vienna: IAEA; 2007. p. 433–48. ISBN 92-0-115506-9.

[52] Sato K. Potential contributions of fast reactor cycle technologies to TRUelement management in Japan. In: Proceedings of technical meeting on fissilematerial management strategies for sustainable nuclear energy. STI/PUB/1288.Vienna: IAEA; 2007. p. 695–725. ISBN 92-0-115506-9.

[53] Salvatores M, Meyer M, Romanello V, Boucher L, Schwenk-Ferrero A. Nuclearfuel cycle synergies and regional scenarios for Europe. Paris: NEA; 2009. ISBN92-64-99086-9.

[54] Stanculescu A. IAEA activities in the area of partitioning and transmutation.Nucl Instr Meth Phys Res A 2006;562:614–7.

[55] Maschek W, Stanculescu A, Arien B, et al. Report on intermediate resultsof the IAEA CRP on ‘studies of advanced reactor technology options foreffective incineration of radioactive waste’. Energy Convers Manage2008;49:1810–9.

[56] Nuclear Energy Agency. Advanced nuclear fuel cycles and radioactive wastemanagement. Paris: NEA; 2006. ISBN: 92-64-02485-9.

1780 J.G. Marques / Energy Conversion and Management 51 (2010) 1774–1780

[57] Nuclear Energy Agency. Management of recyclable fissile and fertile materials.Paris: NEA; 2007. ISBN: 92-64-03255-2.

[58] Nuclear Energy Agency. Strategic and policy issues raised by thetransition from thermal to fast nuclear systems. Paris: NEA; 2009. ISBN:92-64-06064-7.

[59] Martinez-Val J. PATEROS P&T roadmap proposal for advanced fuel cyclesleading to a sustainable nuclear energy: synthesis report. EuropeanCommission FP6 contract FI6W-03418; December 2008. <http://www.sckcen.be/pateros/>.

[60] Mueller AC. Prospects for transmutation of nuclear waste and associatedproton-accelerator technology. Eur Phys J Special Topics 2009;176:179–91.

[61] Abbondanno U, Aerts G, Alvarez H, et al. CERN n_TOF facility: performancereport. CERN-SL-2002-053 ECT. Geneva: CERN; 2002.

[62] US Department of energy research advisory committee and the generation IVinternational forum. A technology roadmap for generation IV nuclear energysystems. GIF-002-00. DOE, GIF; 2002. <http://gif.inel.gov/>.

[63] Vitart X, Le Duigou A, Carles P. Hydrogen production using the sulfur–iodinecycle coupled to a VHTR: an overview. Energy Convers Manage2006;47:2740–7.

[64] Olander D. Nuclear fuels – present and future. J Nucl Mater 2009;389:1–22.[65] Nickel H, Nabielek H, Pott G, Mehner AW. Long time experience with the

development of HTR fuel elements in Germany. Nucl Eng Des2002;217:141–51.

[66] Zimmerman S. Submarine technology for the 21st century, 2nd ed. Victoria:Trafford; 2000. p. 13–33. ISBN 1-55-212330-8.

[67] Haubenreich PN, Engel JR. Experience with the molten-salt reactorexperiment. Nucl Appl Tech 1970;8:118–36.

[68] Generation IV international forum. Annual report 2008. <http://www.gen-4.org/>.

[69] Salvatores M, Khalil H, Bignan G, Hill R, Jacqmin R, Tommasi J. Advanced fastreactor development requirements: is there any need for HEU? Nucl Eng Des2007;237:814–22.

[70] Petti D, Crawford D, Chauvin N. Fuels for advanced nuclear energy systems.MRS Bull 2009;34:40–5.

[71] Comissariat à L’Énergie Atomique. Le Traitement-Recyclage du CombustibleNucléaire Usé. Paris: CEA Saclay and Groupe Moniteur; 2008. p. 155–6. ISBN 2-281-11376-1.

[72] Slessarev I. Intrinsically secure fast reactors with dense cores. Ann Nucl Energy2007;34:883–95.

[73] Guérin Y, Was GS, Zinkle SJ. Materials challenges for advanced nuclear energysystems. MRS Bull 2009;34:10–4.

[74] Abram T, Ion S. Generation-IV nuclear power: a review of the state of thescience. Energy Policy 2008;36:4323–30.

[75] Samaras M, Hoffelner W, Victoria M. Modelling of advanced structuralmaterials for gen IV reactors. J Nucl Mater 2007;371:28–36.

[76] Yvon P, Carré F. Structural materials challenges for advanced reactor systems. JNucl Mater 2009;385:217–22.

[77] European Commission. The sustainable nuclear energy technology platform: avision report. EUR 22842. Luxembourg: Publications Office of the EuropeanUnion; 2007. ISBN 92-79-05591-1.