Chapter 1Nuclear Plants

Abstract Structural materials are important for a wide range of nuclear powerplants. Although the overwhelming majority of current nuclear power plants arelight water reactors advanced plants like Generation IV or fusion are considered asfuture nuclear power options. Current nuclear plants are frequently in the stage oflife-extension programs where damage assessments are most important. Futureplants need predictions of long-term materials behaviour or even new materials tocomply with operation conditions going beyond light water reactors. It is thereforethe aim of this chapter to provide an introduction into operation conditions andmaterials needs of current and future nuclear plants. Changes in policy changepriorities for new plants quickly which can have an impact on priorities discussedin this chapter. The materials issues for the different types of plants remain.

1.1 Current Reactors

The science of atomic radiation and nuclear fission was developed mainly duringthe first half of the last century. During world-war 2 the main interest in nucleartechnology was for the development of the atomic bomb. From 1945 attention wasgiven to this kind of energy for converting it into electricity in safe and reliablenuclear power plants. From the late 1970s to about 2002 the nuclear powerindustry suffered some decline and stagnation. Few new reactors were ordered, thenumber coming on line from mid 1980s little more than matched retirements andmany reactor orders from the 1970s were cancelled. A comprehensive descriptionof the development of nuclear energy can be found e.g. in [1]. Since the nuclearaccident occurring as a result of an earthquake and a tsunami 2011 in Fukushima(Japan) concerns about nuclear fission have been rising again.

Different reactor concepts were designed and partly built over the years usingdifferent cooling media and either a thermal or a fast neutron spectrum. The light

W. Hoffelner, Materials for Nuclear Plants,DOI: 10.1007/978-1-4471-2915-8_1, � Springer-Verlag London Limited 2013

1

water cooled boiling water reactors and pressurised water reactors are today themost important types of reactors for commercial electricity production (about80 %). Besides these types the Canadian CANDU reactors, the advanced gasreactors (AGR) in England, the graphite moderated RMBK in Russia and a fewothers are in operation (see Table 1.1 [2]). A pressurized water reactor (PWR)(Fig. 1.1) uses pressurized (liquid) water as coolant and enriched UO2 tablets incladdings as fuel elements. Extensive public Information about PWRs exist (e.g.[2–4]) and here only the most important and materials related facts shall bementioned.

1.1.1 Pressurized Water Reactors

Pressurized water reactors (PWRs) use ordinary water as both coolant and mod-erator. There are a primary cooling circuit which flows through the core of thereactor under very high pressure, and a secondary circuit in which steam is gen-erated to drive the turbine. A PWR has fuel assemblies of 200–300 rods each,arranged vertically in the core, and a large reactor would have about 150–250 fuelassemblies with 80–100 tonnes of uranium. Water in the reactor core reachesabout 325 �C. It must be kept under about 150 times atmospheric pressure toprevent its boiling. Pressure is maintained by steam in a pressuriser. In the primarycooling circuit the water is also the moderator, and if any of it turned to steam thefission reaction would slow down. This negative feedback effect is one of thesafety features of the type. The secondary shutdown system involves adding boronto the primary circuit. The secondary circuit is under less pressure and the waterhere boils in the heat exchangers which are thus steam generators. The steamdrives the turbine to produce electricity. The unused steam is exhausted into thecondenser where it condenses into water. The resulting water is pumped out of thecondenser with a series of pumps, reheated and pumped back to the reactor vessel.In Russia such PWRs are known as VVER types—water-moderated and—cooled[5, 6]. Under the designation WWER (water water energy reactor) certain types aresummarized by pressurized water reactors of Soviet design. One differentiatesbetween reactors from five generations. The first number indicates usually theapproximate achievement of the reactors, the second number the project name.

1.1.2 Boiling Water Reactors

Also BWRs are well described in the open literature (e.g. [2, 7, 8]).The design of a boiling water reactor (BWR) (Fig. 1.2) has many similarities to

the PWR, except that there is only a single circuit in which the water is at lowerpressure (about 75 times atmospheric pressure) so that it boils in the core at about285 �C. The reactor is designed to operate with 12–15 % of the water in the top

2 1 Nuclear Plants

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1.1 Current Reactors 3

part of the core as steam, and hence with less moderating effect and thus efficiencythere. The steam passes through drier plates (steam separators) above the core andthen directly to the turbines, which are thus part of the reactor circuit. Since thewater around the core of a reactor is always contaminated with traces of radio-nuclides, it means that the turbine must be shielded and radiological protectionprovided during maintenance. Most of the radioactivity in the water is very short-lived (mostly N-16, with a 7 s half-life) so the turbine hall can be entered soonafter the reactor is shut down. The Russian RBMK-reactors are graphite moderatedwater reactors (Fig. 1.3). The RBMK reactor has a huge graphite block structure asthe moderator that slows down the neutrons produced by fission. The graphite

Fig. 1.1 Pressurized water reactor; In a typical commercial pressurized light-water reactor 1 thecore inside the reactor vessel creates heat, 2 pressurized water in the primary coolant loop carriesthe heat to the steam generator, 3 inside the steam generator, heat from the steam, and 4 the steamline directs the steam to the main turbine, causing it to turn the turbine generator, which produceselectricity. (Source US-NRC [4])

4 1 Nuclear Plants

structure is contained in a steel vessel. A helium-nitrogen mixture is used toimprove heat transfer from the graphite to the coolant channels and reduce like-lihood of graphite oxidation. In the RBMK design, boiling occurs. The steamproduced passes to the steam separator for separation of water from the steam. Thesteam then passes to the turbine as in the boiling water reactor design. Similar tothe BWR case, the steam is radioactive, however, the steam separator introduces adelay time so radiation levels near the turbine may not be as high as in the BWRcase.

In contrast to a PWR the RBMK reactor design used at Chernobyl, which usesgraphite instead of water as the moderator and uses boiling water as the coolant,has a large positive thermal coefficient of reactivity, that increases heat generation

Fig. 1.2 Boiling water reactor (BWR); in a typical commercial boiling-water reactor, 1 the coreinside the reactor vessel creates heat, 2 a steam-water mixture is produced when very pure water(reactor coolant) moves upward through the core, absorbing heat, 3 the steam-water mixtureleaves the top of the core and enters the two stages of moisture separation where water dropletsare removed before the steam is allowed to enter the steam line, and 4 the steam line directs thesteam to the main turbine, causing it to turn the turbine generator, which produceselectricity.(Source: USNRC [6])

1.1 Current Reactors 5

when coolant water temperatures increase. This makes the RBMK design lessstable than pressurized water reactors. In addition to its property of slowing downneutrons when serving as a moderator, water also has a property of absorbingneutrons, albeit to a lesser degree. When the coolant water temperature increases,the boiling increases, which creates voids. Thus there is less water to absorbthermal neutrons that have already been slowed down by the graphite moderator,causing an increase in reactivity. This property is called the void coefficient ofreactivity, and in an RBMK reactor like Chernobyl, the void coefficient is positive,and fairly large, causing rapid transients. This design characteristic of the RBMKreactor is generally seen as one of several causes of the Chernobyl accident.

1.1.3 CANDU Reactors

Another type of a water cooled reactor is the CANDU-reactor (Fig. 1.4). CANDU[10, 11] stands for: CANada Deuterium Uranium. The main difference betweenCANDUs and other water moderated reactors is that CANDUs use heavy water forneutron moderation and that they have no pressure vessel. The heavy water sur-rounds the fuel assemblies and primary coolant. The heavy water is unpressurized,

Fig. 1.3 Russian RBMK-reactor; In the RBMK design, boiling occurs. The steam producedpasses to the steam separator which separates water from the steam. The steam then passes to theturbine as in the boiling water reactor design. (Source http://en.wikipedia.org/wiki/File:RBMK_reactor_schematic.svg, [9])

6 1 Nuclear Plants

and a cooling system is required to keep it from boiling. Instead in a pressurevessel, the pressure is contained in much smaller tubes that contain the fuelbundles. These smaller tubes are easier to fabricate than a large pressure vessel.They are made of a zirconium alloy (zirconium ? 2.5 % wt niobium), similar tofuel claddings in LWRs. The zircaloy tubes are surrounded by a much larger low-pressure tank known as a calandria, which contains the majority of the moderator.The CANDU was designed to use natural uranium as its fuel. Traditional designsusing light water as a moderator will absorb too many neutrons to allow a chainreaction to occur in natural uranium due to the low density of active nuclei. Heavywater absorbs fewer neutrons than light water, allowing a high neutron economythat can sustain a chain reaction even in unenriched fuel. Also, the low temperatureof the moderator (below the boiling point of water) reduces changes in the neu-trons’ speeds from collisions with the moving particles of the moderator (‘‘neutronscattering’’). The neutrons therefore are easier to keep near the optimum speed tocause fissioning; they have good spectral purity. At the same time, they are stillsomewhat scattered, giving an efficient range of neutron energies. The largethermal mass of the moderator provides a significant heat sink that acts as anadditional safety feature. If a fuel assembly were to overheat and deform within itsfuel channel, the resulting change of geometry permits high heat transfer to thecool moderator, thus preventing the breach of the fuel channel, and the possibilityof a meltdown. Furthermore, because of the use of natural uranium as fuel, thisreactor cannot sustain a chain reaction if its original fuel channel geometry isaltered in any significant manner.

Fig. 1.4 Schematic drawing of a CANDU reactor; 1 fuel bundle, 2 reactor core (calandria),3 control rods, 4 heavy water reservoir, 5 steam generator, 6 light water pump, 7 heavy waterpump, 8 refuelling device, 9 heavy water moderator, 10 pressure tube, 11 to steam turbine, 12from steam turbine 13 containment building. (Source Wikipedia [8])

1.1 Current Reactors 7

In a traditional light water reactor (LWR) design, the entire reactor core is asingle large pressure vessel containing the light water, which acts as moderator andcoolant, and the fuel arranged in a series of long bundles running the length of thecore. To refuel such a reactor, it must be shut down, the pressure dropped, the lidremoved, and a significant fraction of the core inventory, such as one-third,replaced in a batch procedure. The CANDU‘s calandria-based design allowsindividual fuel bundles to be removed without taking the reactor off-line.A CANDU fuel assembly consists of a number of zircaloy tubes containingceramic pellets of fuel arranged into a cylinder that fits within the fuel channel inthe reactor.

1.1.4 Advanced Gas Reactors

The last type of current moderated reactors which will be introduced using athermal neutron spectrum is the British advanced gas reactor (AGR) [12] shown inFig. 1.5. At the heart of the reactor is a graphite core called the moderator.Running vertically through this core are tubes containing uranium called fuelchannels. The moderator has a vital role to play as it slows down the neutronsreleased by the fuel so that they will interact with other uranium atoms and sustain

Fig. 1.5 Advanced gas reactor (AGR); the heat exchanger is contained within the steel-reinforced concrete combined pressure vessel and radiation shield. 1 Charge tubes, 2 Controlrods, 3 Graphite moderator, 4 Fuel assemblies, 5 Concrete pressure vessel and radiation shielding,6 Gas circulator (CO2), 7 Water, 8 Water circulator, 9 Heat exchanger, 10 Steam, (Source: http://en.wikipedia.org/wiki/File:AGR_reactor_schematic.svg)

8 1 Nuclear Plants

the chain reaction. The coolant is CO2. The AGR was developed to operate at ahigher gas temperature for improved thermal efficiency, requiring stainless steelfuel cladding to withstand the higher temperature. Because the stainless steel fuelcladding has a higher neutron capture cross section than former Magnox fuel cans,enriched uranium fuel is needed, with the benefit of higher ‘‘burn ups’’ of18,000 MWt-days per tonne of fuel, requiring less frequent refuelling.

1.2 Improvements and Developments of Reactor Concepts

For the nearer future traditional light or heavy water reactors will be the choice oftechnology. Also this group underwent and undergoes significant improvements insafety and in performance. Besides traditional large nuclear power stations alsosmall reactors for local energy supply (electric and thermal) are studied in differentcountries. It may therefore be useful to consider also these reactor developments inthis section in addition to the current stage of the generation IV projects. Lightwater reactors are still considered as the major plants for future energy supply. Theconcepts for the next generation LWR are: (1) Highest safety and economy in the2030 timeframe (2) Simplifying operation and maintenance (3) Dramaticallyshorten time for construction, (4) Reduction of quantity of spent fuel produced,reduction of consumption of uranium, radioactive waste and exposure to radiation,and (5) Improvement of performance for plant life (approximately 80 years).

1.2.1 Advanced Light Water Reactors

Light water reactors currently offered represent the so called 3rd generation ofnuclear power plants as discussed later in more detail. In the following the mostimportant types and projects on advanced light water reactors are summarized.

This summary follows a more detailed description given in [13]. Comparedwith current LWRs advanced plants are considered to have:

• a standardised design for each type to expedite licensing, reduce capital cost andreduce construction time,

• a simpler and more rugged design, making them easier to operate and lessvulnerable to operational upsets,

• higher availability and longer operating life—typically 60 years,• further reduced possibility of core melt accidents,• resistance to serious damage that would allow radiological release from an

aircraft impact,• higher burn-up to reduce fuel use and the amount of waste,• burnable absorbers (‘‘poisons’’) to extend fuel life.

1.1 Current Reactors 9

The greatest departure from second-generation designs are many incorporatepassive or inherent safety features which require no active controls or operationalintervention to avoid accidents in the event of malfunction. They may rely ongravity, natural convection or resistance to high temperatures. Some details con-cerning advanced PWRs and BWRs are given in Tables 1.2 and 1.3. The importanttypes of 3rd generation plants are:

Advanced boiling water reactor (ABWR) derived from a General Electricdesign.

System 80+, is an advanced pressurised water reactor (PWR), which was readyfor commercialisation but is not now being promoted for sale.

The Westinghouse AP1000, scaled-up from the AP600, received final designcertification from the NRC in December 2005—the first Generation 3 ? type to doso. It represents the culmination of a 1,300 man-year and $440 million design andtesting program.

GE Hitachi Nuclear Energy’s ESBWR is a Generation III ? technology thatutilizes passive safety features and natural circulation principles and is essentiallyan evolution from its predecessor design, the SBWR at 670 MWe.

Mitsubishi’s large APWR (1,538 MWe)—advanced PWR—was developed incollaboration with four utilities (Westinghouse was earlier involved).

Areva NP (formerly Framatome ANP) has developed a large (1,600 and up to1,750 MWe) European pressurised water reactor (EPR), which is currently underconstruction in Finnland.

Table 1.3 Typical data of an advanced BWR [14, 15]

Electric output 1,700–1,800 MWeFuel Large bundleSafety system Hybride (optimizes passive and active safety)Primary containment vessel Double containments

Outer: Steel containment vesselInner: Steel plate reinforced concrete containment vessel

Countermeasures of external events Seismic isolation systemsReinforced buildingEarthquake

Aiplane crash

Table 1.2 Typical data of an advanced PWR (GWd/t means GW-days per ton uranium) [14, 15]

Electric output 1,780 MWe (Plant efficiency 40 %)Fuel average burn-up [70 GWd/tPrimary coolant temperature (Hot Leg) 330 �CSteam generator surface area 8,500 m3 (for High efficiency)Primary coolant flow rate 29,000 m3/h/loopSafety system 4 Train direct-air-cooling hybrid systemUltimate heat sink Air and sea water

10 1 Nuclear Plants

Together with German utilities and safety authorities, Areva NP (FramatomeANP) is also developing another evolutionary design, the SWR 1,000, a1,250 MWe BWR with 60 year design life now known as Kerena.

Toshiba has been developing its evolutionary advanced BWR (1,500 MWe)design, originally BWR 90+ from ABB then Westinghouse, working with Scan-dinavian utilities to meet European requirements.

A third-generation standardised VVER-1200 reactor of 1,150–1,200 MWe is,amongst others, an evolutionary development of the well-proven VVER-1000 inRussia.

1.2.2 Advanced Heavy Water Reactors

The CANDU-9 (925–1,300 MWe) was developed also as a single-unit plant. It hasflexible fuel requirements ranging from natural uranium through slightly-enricheduranium, recovered uranium from reprocessing spent PWR fuel, mixed oxide(U and Pu) fuel, direct use of spent PWR fuel, to thorium. India is developing theadvanced heavy water reactor (AHWR) as the third stage in its plan to utilisethorium to fuel its overall nuclear power program. The AHWR is a 300 MWe

reactor moderated by heavy water at low pressure.

1.2.3 Small Modular Reactors

As nuclear power generation has become established since the 1959s, the size ofthe reactor units has grown from 60 MWe to more than 1,600 MWe, with cor-responding economies of scale in operation. At the same time there have beenmany hundreds of smaller reactors built both for naval use (up to 190 MWthermal) and as neutron sources, yielding enormous expertise in the engineering ofsmall units. The international atomic energy agency (IAEA) defines ‘small’ asunder 300 MWe. The contents of this subsection are based on information given in[16] and [17].

Designs for SMRs are being developed in several countries, often throughcooperation between government and industry. Countries involved includeArgentina, China, Japan, Korea, Russia, South Africa and the United States. SMRdesigns encompass a range of technologies, some being variants of the six Gen-eration IV systems selected by GIF, while others are based on established LWRtechnology.

Such reactors could be deployed as single or double units in remote areaswithout strong grid systems, or to provide small capacity increments on multi-unitsites in larger grids. They feature simplified designs and would be mainly factory-fabricated, potentially offering lower costs for serial production. Their much lower

1.2 Improvements and Developments of Reactor Concepts 11

capital cost and faster construction than large nuclear units should make financingeasier. Other advantages could be in the area of proliferation resistance, as somedesigns would require no on-site refueling, while others would only requirerefueling after several years. Some could be used with advanced fuel cycles,burning recycled materials.

Numerous concepts exist for SMRs based on LWR technology. Several suchdesigns are being promoted by nuclear industry companies, including AREVA,Babcock & Wilcox (mPower), General Atomics, NuScale and Westinghouse(IRIS). Others are being developed by national research institutes in Argentina,China, Japan, Korea and Russia. Two small units designed to supply electricityand heat are under construction in Russia, based on existing ice-breaker propulsionreactors; these will be barge-mounted for deployment to a remote coastal settle-ment on the Kamchatka peninsula. Some other designs are well advanced withinitial licensing activities underway.

Some SMR designs are HTRs. These designs are well suited to heat orco-generation applications as discussed later. There are also several other conceptsfor advanced SMR designs, including liquid metal-cooled fast reactors. They aregenerally at an earlier stage of development, and are subject of GIF collaborativeefforts. One example in this category is the 4S design from Toshiba of Japan, asodium-cooled ‘‘nuclear battery’’ system capable of operating for 30 years with norefueling. It has been proposed to build the first such plant to provide 10 MW ofelectricity to a remote settlement in Alaska, and initial licensing procedures havebegun. Another example in this category is the Hyperion Power Module, a lead–bismuth cooled LMR, developed by Hyperion Power. Other concepts for advancedSMRs have been proposed by commercial and research organizations in severalcountries, and some aim to commence licensing activities in the next few years.

A recent SMR candidate is the travelling wave reactor, which is currentlypromoted by TerraPower [18]. According to [19], a traveling-wave reactorrequires very little enriched uranium, reducing the risk of weapons proliferation.The reactor uses depleted-uranium fuel packed inside hundreds of hexagonalpillars. In a ‘‘wave’’ that moves through the core at only a centimeter per year, thisfuel is transformed (or bred) into plutonium, which then undergoes fission. Thereaction requires a small amount of enriched uranium to get started and could runfor decades without refueling. The reactor uses liquid sodium as a coolant; coretemperatures are rather hot–about 550 �C, versus the 330 �C typical of conven-tional reactors.

If multiple SMR units on a single site were to become a competitive alternativeto building one or two large units, then SMRs could eventually form a significantcomponent of nuclear capacity. They could also enable the use of nuclear energyin locations unsuitable for large units, and some designs could extend its use fornon-electricity applications. However, whether SMR designs can be successfullycommercialized, with an overall cost per unit of electricity produced that iscompetitive with larger nuclear power plants and other generating options, remainsto be seen.

12 1 Nuclear Plants

1.2.4 Advanced New Reactor Concepts

In the new century several factors have combined to revive the prospects fornuclear power. First is realisation of the scale of projected increased electricitydemand worldwide, but particularly in rapidly-developing countries. Secondly isawareness of the importance of energy security, and thirdly is the need to limitcarbon emissions due to concern about global warming. With minimal greenhousegas emissions, nuclear energy can safely provide the world with not only electricalenergy production but also process-heat energy production. Examples of thebenefits that can be derived from process heat generation include the generation ofhydrogen, the production of steam for extraction of oil-in-oil sand deposits, and theproduction of process heat for other industries so that natural gas or oil do not haveto be used. In 1999, an international collaborative initiative for the development ofadvanced (Generation IV) reactors was started [20]. The idea behind this effort wasto bring nuclear energy closer to the needs of sustainability, to increase prolifer-ation resistance and to support concepts able to produce energy (both electricityand process heat) at competitive costs (see Table 1.2). Six reactor concepts werechosen for further development:

• sodium fast reactor (SFR)• very high temperature gas-cooled reactor (VHTR)• lead or lead–bismuth cooled liquid metal reactor (LMR)• helium gas-cooled fast reactor (GFR)• molten salt reactor (MSR)• supercritical water reactor (SCWR)

In view of sustainability, the Generation IV reactors should not only havesuperior fuel cycles to minimize nuclear waste, but they should also be able toproduce process heat or steam for hydrogen-production, synthetic fuels, refineryprocesses and other commercial uses. These reactor types were described in the2002 Generation IV roadmap. Different projects around the world have beenstarted since that time. The most advanced efforts are in reactors where productionexperience already existed. These reactors are the SFR and the VHTR. The otherreactor types are still more in a design concept phase. These new technologieshave also created remarkable demands on materials compared to LWRs. Highertemperatures, higher neutron doses, environments very different from water anddesign lives of 60 years present a real engineering challenge.

The development of nuclear power can be divided into several plant-generationsas shown in Fig. 1.6. Advanced reactors, based on current nuclear power planttechnology (EPR, AP1000, ESBWR, advanced CANDU, APWR etc.) are calledGeneration III+. Generation IV reactors go beyond LWR-technology. Although theywere built in the past at least on demonstration level, they are intended to be com-mercially available along the guidelines given in Table 1.4 by about 2030. It wasalso recognized that joint international R&D would be necessary to meet thisambitious goal. Six concepts as mentioned before were chosen for further R&D

1.2 Improvements and Developments of Reactor Concepts 13

work. Fusion reactors are sometimes called Generation V nuclear power plants.According to [20] the R&D performed within GIF focuses on both the viability andperformance phases of system development: The former phase examines the fea-sibility of key technologies, such as, for example, suitable or novel structuralmaterials or advanced fuel concepts. The latter phase focuses on the development ofperformance data and optimization of the system. The original scope of GIFactivities did not contain also a demonstration phase, which involves the detaileddesign, licensing, construction and operation of a prototype or demonstration systemin partnership with industry. However, in current projects a closer relation betweenGIF-projects and demonstrators becomes visible. Other international collaborationsin the field of advanced reactors exist also. They are, however rather complementaryto GIF than competing with GIF. They are described here following mainly [15].One of them is the former global nuclear energy partnership (GNEP). The globalnuclear energy partnership (GNEP) was originally formed to control the interna-tional fuel cycle and to avoid proliferation risks. In 2010 its name was changed tointernational framework for nuclear energy cooperation (IFNEC) [21] and a newmission statement was established with the aims to broaden the scope to widerinternational participation to accelerate development and deployment of advancedfuel cycle technologies to encourage clean development and prosperity worldwide,improve the environment, and reduce the risk of nuclear proliferation. The DOE hasoutlined four overarching goals for the IFNEC: (1) to decrease U.S. reliance onforeign energy sources without impeding U.S. economic growth; (2) to employimproved technologies to recover more energy and reduce waste when recyclingspent nuclear fuel; (3) to encourage the use of energy sources that emit the leastatmospheric greenhouse gases; and (4) to reduce the threat of nuclear proliferation.The partnership has a three-tiered organization structure. The Executive Committee

Fig. 1.6 Development path of nuclear power plants [20]

14 1 Nuclear Plants

comprised of Ministerial-level officials, provides the high-level direction. TheSteering Group, whose members are designated by the Executive Committee, car-ries out actions on behalf of IFNEC at the direction of the Executive Committee. At aSeptember 2007 meeting of the Executive Committee, two working groups wereestablished to address matters concerning ‘reliable nuclear fuel services’ and‘infrastructure development’. Currently, the Nuclear Fuel Service Working Group isaddressing how to design and implement an effective nuclear energy infrastructureemploying fuel leasing and other economically viable and proliferation-securearrangements. The Infrastructure Development Working Group is addressing thefinancial, technical, and human resource issues involved in creating an internationalnuclear energy architecture based on IFNEC’s Statement of Principles. In October2007, the DOE announced the first set of technical and conceptual design devel-opment awards—over $16.3 million to four multinational industry consortia led byAreva, Energy Solutions, GE-Hitachi Nuclear Americas, and General Atomics.In announcing the decision, Assistant Secretary of Nuclear Energy said that thegrants ‘‘enable DOE to benefit from the vast technological and business experienceof the private sector as we move towards the goal of closing the nuclear fuel cycle’’.In a statement April 2009, the DOE announced that the Department has cancelled theUS domestic component of the IFNEC [22]. It further said, ‘‘The long-term fuelcycle research and development program will continue but not the near-term

Table 1.4 Goals for Generation IV nuclear power plants as defined by the international gen-eration IV initiative (GIF) [20]

Goals for generation IV nuclear systems

Sustainability Generation IV nuclear energy systems will provide sustainableenergy generation that meets clean air objectives andpromotes long-term availability of systems and effective fuelutilization for worldwide energy production

Generation IV nuclear energy systems will minimize andmanage their nuclear waste and notably reduce the long-term stewardship burden, thereby improving protection forthe public health and the environment

Economics Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources

Generation IV nuclear energy systems will have a level offinancial risk comparable to other energy projects

Safety and reliability Generation IV nuclear energy systems operations will excel insafety and reliability

Generation IV nuclear energy systems will have a very lowlikelihood and degree of reactor core damage

Generation IV nuclear energy systems will eliminate the needfor offsite emergency response

Proliferation resistance andphysical protection

Generation IV nuclear energy systems will increase theassurance that they are a very unattractive and the leastdesirable route for diversion or theft of weapons-usablematerials, and provide increased physical protection againstacts of terrorism

1.2 Improvements and Developments of Reactor Concepts 15

deployment of recycling facilities or fast reactors. The international component ofGNEP is under interagency review’’.

The international atomic energy agency (IAEA) project INPRO was establishedin 2001 by bringing together technology holders, users, and potential users toconsider jointly the international and national actions required for achievingdesired innovations in nuclear reactors and fuel cycles [23, 24]. Since the earlypart of 2009, it has been determined to structure the project’s task into the fol-lowing four areas, with a forum for dialogue by members as a crosscutting vehiclefor communication:

• Methodology development and its use by members,• Future nuclear energy vision and scenario,• Innovative technologies,• Innovation in institutional arrangement.

The first results of the INPRO activity are listed in Ref. [25] for the assessmentof innovative nuclear reactors and fuel cycles.

1.3 Neutron Spectrum, Fast Reactors and Fuel Cycles

1.3.1 Neutron Spectrum

Before proceeding further with description of advanced reactors its possiblerelation to advanced fuel cycles shall be highlighted. Almost all of the six GenIVplants are fast reactors operating—in contrast to current LWRs—without moder-ator. Such fast breeder reactors have already been in operation but many of themwere shut down or they went never into operation on a commercial scale fordifferent reasons (see sodium fast reactors later in this chapter). The spectrum ofneutron energies produced by fission varies significantly from the energy spec-trum, or flux, existing in a moderated reactor. Figure 1.7 [26] illustrates thedifference in neutron flux spectra between a thermal reactor and a fast breederreactor. The energy distribution of neutrons from fission is essentially the same forboth reactors, so the differences in the curve shapes may be attributed to theneutron moderation or slowing down effects. No attempt is made to thermalize orslow down neutrons in the fast breeder reactor (e. g. liquid metal cooled); there-fore, an insignificant number of neutrons exist in the thermal range. For thethermal reactor (water moderated), the spectrum of neutrons in the fast region([0.1 MeV) has a shape similar to that for the spectrum of neutrons emitted by thefission process. In the thermal reactor, the flux in the intermediate energy region(1 eV to 0.1 MeV) has approximately a 1/E dependence which is caused by theslowing down process, where elastic collisions remove a constant fraction ofthe neutron energy per collision (on the average), independent of energy; thus, theneutron loses larger amounts of energy per collision at higher energies than at

16 1 Nuclear Plants

lower energies. The fact that the neutrons lose a constant fraction of energy percollision causes the neutrons to tend to ‘‘pile up’’ at lower energies, that is,a greater number of neutrons exist at the lower energies as a result of this behavior.Following up [27] one can say that a fast reactor is a category of nuclear reactors inwhich the fission chain reaction is sustained by fast neutrons. Such a reactor needsno neutron moderator, but must use fuel that is relatively rich in fissile materialwhen compared to that required for a thermal reactor.

On average, more neutrons per fission are produced from fissions caused by fastneutrons than from those caused by thermal neutrons. This results in a largersurplus of neutrons beyond those required to sustain the chain reaction. Theseneutrons can be used to produce extra fuel, or to transmute long-halflife waste toless troublesome isotopes, such as was done at the Phénix reactor in Marcoule inFrance, or some can be used for each purpose. Though conventional thermalreactors also produce excess neutrons, fast reactors can produce enough of them tobreed more fuel than they consume. Such designs are known as fast breederreactors.

Fast neutrons also have an advantage in the transmutation of nuclear waste. Thereason for this is that the ratio between the fission cross section and the absorptioncross section of a plutonium or minor actinide nuclide is often higher in a fastspectrum than in a thermal or epithermal spectrum.

In practice sustaining a fission chain reaction with fast neutrons means usingrelatively highly enriched uranium or plutonium. The reason for this is that fissilereactions are favored at thermal energies, since the ratio between the Pu-239fission cross section and U-238 absorption cross section is *100 in a thermal

Fig. 1.7 Comparison of neutron flux spectra for termal and fast breeder reactor (Source [26])

1.3 Neutron Spectrum, Fast Reactors and Fuel Cycles 17

spectrum and 8 in a fast spectrum. Therefore it is impossible to build a fast reactorusing only natural uranium fuel. However, it is possible to build a fast reactor thatwill breed fuel (from fertile material) by producing more fissile material than itconsumes. After the initial fuel charge such a reactor can be refueled by repro-cessing. Fission products can be replaced by adding natural or even depleteduranium with no further enrichment required. This is the concept of the fastbreeder reactor or FBR.

1.3.2 Fuel Cycles

1.3.2.1 Uranium/Plutonium Based Fuel Cycles

Uranium resources and nuclear waste were mentioned already in the introductionas driving forces for further developments of nuclear energy. This has an impacton the selection of advanced reactor concepts. Figure 1.8 compares different fuelcycles and its consequences.

• Once through cycle• Limited recycle• Full recycle

In a once-through fuel cycle the spent fuel consisting of plutonium, uranium,neptunium, minor actinides (ameritium, curium) and fission products is disposed

Fig. 1.8 Possible nuclear fuel management options (after [28])

18 1 Nuclear Plants

of in a final repository. In case of fuel-reprocessing uranium and plutonium areseparated. Only the still usable portion of uranium is recycled and the rest isdisposed of. Separation can be done either chemically (liquid extraction) orelectro-metallurgically. High amounts of uranium are still lost this way and plu-tonium together with the minor actinides are the long-living elements in thenuclear waste.

Additionally, plutonium bears a high proliferation risk. The full recycle optionuses the fact that fast reactors can operate with mixed fuel containing uranium,plutonium and minor actinides which allows fuel cycles where only the fissionproducts remain in the waste to be disposed of. As they have much shorter life-time than the plutonium and the actinides, the life-time of the waste becomes muchshorter (see Fig. 1.9). Also uranium from the waste can be re-used. This meansthat the uranium resources last for very long period of time and nuclear wastewould no longer contain long-living products. This full recycle option is furtherillustrated in Fig. 1.10. Basically there are two routes for fuel treatment consid-ered: (1) To separate U and Pu (as already done) but to separate also the minoractinides and to produce mixed fuel. Weapon-grade plutonium remains separate inthis process chain until mixing which is considered as a proliferation risk.Therefore concepts are under development to separate uranium, plutonium, nep-tunium and the minor actinides in one step where plutonium does not appear as aseparate fraction. The two concepts are summarized in Fig. 1.10. The fuel- and

Fig. 1.9 Influence of advanced fuel cycle on life-time and radio-toxicity of high level waste(ALI: annual limit on intake) (Source [29])

1.3 Neutron Spectrum, Fast Reactors and Fuel Cycles 19

fuel cycle options for the different plants are comprehensively described for thedifferent reactor types in [33]. The global actinide cycle international demon-stration (GACID) [30] project shall demonstrate that the SFR can manage effec-tively all actinide elements in the fuel cycle, including uranium, plutonium, andminor actinides (neptunium, americium and curium). Plans for an advancedrecycling center (ARC) of GE Hitachi are well in progress [31] The ARC startswith the separations of spent nuclear fuel into three components: (1) uranium thatcan be used in CANDU reactors or re-enriched for use in LWRs; (2) fissionproducts (with a shorter half life) that are stabilized in glass or metallic form forgeologic disposal; and (3) actinides (the long-lived radioactive material in spentnuclear fuel) which are used as fuel in the advanced recycling reactor (ARR).

An electrometallurgical process is proposed to perform separations. Thisprocess uses electric current passing through a salt bath to separate the componentsof spent nuclear fuel. A major advantage of this process is that it is a dry process(the processing materials are solids at room temperature). This significantlyreduces the risk of inadvertent environmental releases. Additionally, unliketraditional aqueous MOX separations technology, electrometallurgical separationsdoes not generate separated pure plutonium making electrometallurgical separa-tions more proliferation resistant. The actinide fuel (including elements such as

Fig. 1.10 Concepts for advanced fuel recycling. Option 1 consists of two aqueous separationsteps where U, Pu and Np are extracted in one stage and the minor actinides are extracted inanother stage. The GANEX process releases U, Pu and the minor actinides in one process step.For both options only the fission products (FP) have to be disposed [29]

20 1 Nuclear Plants

plutonium, americium, neptunium, and curium) manufactured from the separationsstep is then used in PRISM to produce electricity in a conventional steam turbine.Figure 1.11 shows a schematic of the ARC. The sodium coolant in the PRISM or‘burner’ reactor, allows the neutrons to have a higher energy, converting them intoshorter-lived fission products. An ARC is proposed to consist of an electromet-allurgical separations plant and three power blocks of 622 MWe each for a total of1,866 MWe [31].

Besides the well known oxide fuel also other types like carbides, nitrides ormetallic fuel options are currently considered as options.

1.3.2.2 Other Fuel Cycles

Thorium Cycle

Alternatively to uranium/plutonium based fuels thorium fuel cycles are explored tobecome independent from uranium supply. India has envisaged robust thoriumreactor technologies as a promising sustainable future energy resource for thecountry. Studies indicate that once the FBR capacity reaches about 200 GWe,thorium-based fuel can be introduced progressively in the FBRs to initiate the thirdstage of the program, where the U-233 bred in these reactors is to be used in thethorium based reactors [32]. The proposed road map for the third stage thereforecomprises thorium-based reactor technologies, incorporating the (Th-U-233)cycle. India is one of the leading countries in the world in thorium research and hasgained the experience through thorium irradiation and the operation of U-233fuelled research reactors.

Fig. 1.11 Schematic of the advanced recycling center ARC of GE-Hitachi

1.3 Neutron Spectrum, Fast Reactors and Fuel Cycles 21

Molten Salt

A total different type of fuel is used in molten salt reactors. In such reactors thefuel can be dissolved in the coolant which means that fuel and cooling becomethe same medium. Currently, uranium -, plutonium- and thorium [33] based fuelsare considered. A more extended discussion of current molten salt concepts can befound later in this chapter.

1.4 Generation IV Nuclear Plants

The six nuclear technologies proposed within GIF are not entirely new plants.They are based on some experience gained with experimental reactors or evenwith large scale pilot plants like the sodium cooled French Superphenix [34] or thegas cooled German HTR [35]. The SCWR is basically a pressurized light waterreactor running a supercritical steam cycle which has an impact on pressure andtemperature. Most plant experience exists with SFRs and HTRs. This is the reasonwhy we will put the emphasize here on these two types of ractors mainly. Lists ofSFR and HTR plants can be found in the literature e.g. [36, 37]. An assessment ofthe timeline for deployment of several Generation IV nuclear systems is shown inFig. 1.12 [38]. The description of the different types of Generation IV plants willfollow mainly the views described in [15, 20, 39].

The next group behind the most advanced concepts SFR and VHTR are:SCWR, LFR and GFR with expected demonstrator availability of 2025. Veryinteresting, but least developed is the MSR with the longest expected time for ademonstrator. Even if the absolute values of the time scale might be disputableFig. 1.12 gives quite a good picture about the maturity of the different systems.Besides fuel and fuel cycle as outlined above, structural materials are consideredas a key issue for almost several concepts. Performance of components underservice conditions which are different to current light water reactors is also aconsiderable challenge for design and design codes.

1.4.1 Sodium Fast Reactor

1.4.1.1 Technology Base

The Sodium-Cooled Fast Reactor (SFR) system works with a fast-spectrum reactorand closed fuel recycle. The primary mission for the SFR is management of high-level wastes and, in particular, management of plutonium and other actinides.SFRs are not a new development and the history of SFRs including planned plantsare shown in Table 1.5. Insufficient plant availability and high cost were mainproblems why SFR-projects failed. With innovations to reduce capital cost, it is

22 1 Nuclear Plants

expected that future SFRs can extend to electricity production, given the provencapability of sodium reactors to utilize almost all of the energy in the naturaluranium versus the 1 % utilized in thermal spectrum systems. Plant size optionsfor SFRs range from from modular systems of a few hundred MWe to largemonolithic reactors of 1,500–1,700 MWe. Sodium core outlet temperatures aretypically 530–550 �C.

The primary coolant system can either be arranged in a pool layout as shown inFig. 1.13 (a common approach, where all primary system components are housed ina single vessel), or in a compact loop layout, favored in Japan where pool pump andheat exchanger are placed outside of the reactor pool. Both options have a relativelyhigh thermal inertia of the primary coolant. A large margin to coolant boiling isachieved by design, which is an important safety feature of these systems. Anothermajor safety feature is that the primary system operates at essentially atmosphericpressure, pressurized only to the extent needed to move fluid. This avoids thenecessity of a reactor pressure vessel. Sodium reacts chemically with air, and withwater, and thus the design must limit the potential for such reactions and theirconsequences. To improve safety, a secondary sodium system acts as a bufferbetween the radioactive sodium in the primary system and the steam or water that iscontained in the conventional Rankine-cycle power plant. If a sodium-water reac-tion occurs, it does not involve a radioactive release. Two fuel options exist for theSFR: (1) MOX and (2) mixed uranium–plutonium–zirconium metal alloy (metal).The experience with MOX fuel is considerably better than with metallic fuel. SFRsrequire a closed fuel cycle to enable their advantageous actinide management andfuel utilization features described above. The fuel cycle technologies must beadaptable to thermal spectrum fuels in addition to serving the needs of the SFR

Fig. 1.12 The deployment perspectives of advanced nuclear plants (after [38]). Most importantresearch and development activities are also shown

1.4 Generation IV Nuclear Plants 23

because of the following reasons: First, the startup fuel for the fast reactors mustcome ultimately from spent thermal reactor fuel. Second, for the waste managementadvantages of the advanced fuel cycles to be realized fuel from thermal spectrumplants will need to be processed with the same recovery factors. Thus, the reactortechnology and the fuel cycle technology are strongly linked [39].

Fig. 1.13 Schematic of a sodium fast reactor in a pool layout (Source: US-DOE, http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf)

Table 1.5 Status of sodium fast reactors worldwide

U.S. Europe Russia Asia

Past Clementine, EBR1/11, SEFOR,FFTF

Dounreay,Rhapsody,Superphenix

BN-350

Cancelled Clinch River, IFR SNR-300Operating Phenix BN-600 Joyo, FBTR,

MonjuUnder

constructionBN-800 PBFR, CEFR

Planned S4, PRISM ASTRID BN-1800

S4, JSFR,KALIMER

24 1 Nuclear Plants

1.4.1.2 SFRs in Japan

Innovative concepts and technologies for SFRs are shown in Table 1.6 taking theJapanese JSFR as an example (for more details see [40]). The size of a reactorvessel for the JSFR of an advanced loop-type SFR will be minimized and thereactor core internals will be simplified. The diameter and wall thickness of thereactor vessel are considered to be 10.7 m and 50–60 mm, respectively. A short-ened piping, two-loop cooling system and integrated intermediate heat exchanger(IHX) with a primary pump are introduced into the design from the view point ofreduction of cost, safety, maintainability and manufacturability. A containmentvessel would be rectangular in shape, because the pressure load to the vessel is nothigh compared with that to light water reactors. A double-wall structure of steelplate reinforced concrete is applied to all parts of the building. The volume of thereactor building is about 150,000 m3, which is less than one-half of a currentadvanced PWR. Regarding demonstration and commercialization of the JSFR,there are several innovative technologies studied. The current status of innovativetechnologies under development include the two-loop cooling system, increasedreliability of the reactor system, a simplified fuel-handling system, a passivereactor shutdown system, mitigation measures against core disruptive accidents,and a minor actinide (MA)—bearing MOX (U/Pu mixed oxide) fuel core.Economic assessments were also peformed. It was shown that based on: reductionof reactor building volume and structural weight, adopting a simplified configu-ration, and pursuing scale merit by enlargement of the power output, the con-struction cost per unit of electricity for the JSFR would be competitive with that offuture light water reactors [15, 40] (Table 1.6).

1.4.1.3 SFR Projects in Russia

Following recent information from the literature [41] has Russia a long experiencewith sodium cooled reactors. The BN-350 prototype FBR generated power inKazakhstan for 27 years to 1999 and about half of its 1,000 MW (thermal) outputwas used for water desalination. It used uranium enriched to 17–26 %. Its designlife was 20 years, and after 1993 it operated on the basis of annual licence renewal.Russia’s BOR-60 was a demonstration model preceeding it.The Construction ofthe first BN-800 reactor is well advanced. It has improved features including fuelflexibility: U ? Pu nitride, MOX, or metal, and with breeding ratio up to 1.3.However, during the plutonium disposition campaign it will be operated with abreeding ratio of less than one. It has much enhanced safety and improvedeconomy; operating cost is expected to be only 15 % more than VVER. It iscapable of burning up to 2 t of plutonium per year from dismantled weapons andwill test the recycling of minor actinides in the fuel. In 2009 two BN-800 reactorswere sold to China, with construction due to start in 2011.

The BN-1800 is next in this chain. Its power-generating unit is designed to meetthe requirements of the strategy for developing atomic energy in Russia in the first

1.4 Generation IV Nuclear Plants 25

half of the 21st century. The development time is the next 15 years and constructioncould start after 2020. The design includes the development of advanced technicalsolutions as compared with the BN-800 reactor which is now under construction. Thenew technical solutions are based on the substantial positive experience in operatingfast reactors in Russia (*125 reactor�years), specifically the BN-600 reactor. Theinnovations make it possible not only to solve strategic problems, such as increasingsafety, improving ecology (by burning actinides), and nonproliferation but also tomake large improvements in economic performance. The development of BN-1800is based on the maximum possible use of tested solutions, implemented in BN-350,-600, and -800 reactors and the use of new technical solutions which increasesafety and cost-effectiveness. The following technical solutions have been tested:

• three-loop scheme for the power-generating unit, sodium in the first and secondloops, working body water/steam;

• integrated arrangement of the first (radioactive) loop with the main and backupvessels.

Economic performance is improved by the following:

• increasing the power• increasing the efficiency of the steam-power cycle up to 45.5–47 % by increasing

the coolant temperature in the three loops, using the working body in the thirdloop with transcritical pressure, using schemes with intermediate superheating ofsteam, and optimizing the construction and layout of the turbine system;

• increasing the rated service life of the power-generating unit up to 60 year,increasing the service life of the replaceable equipment by a factor of 1.5–2compared with that achieved in BN-600.

Table 1.6 Innovative concepts and technologies for SFRs taking the Japanese JSFR as anexample, IHX…intermediate heat exchanger, CV…core vessel, SG…steam generator (after [40])

Economy Higher reliabilityReduction of mass and volume Sodium technology• Shortened piping with high chromium steel • Sodium leak tightness with double wall piping• Two loop cooling system • Higher reliable SG with double wall tube• Integrated pump-IHX component • Higher maintenance ability inside of sodium

boundary• Compact reactor vessel• Simplified fuel handling system• CV with steel plate reinforced concrete

buildingLong operation by high burn-up fuel Higher safety• Advanced fuel materials

Core safety• Passive shutdown and decay heat removal• Re-criticality free coreSeismic reliability• Seismic reliability in core assemblies

26 1 Nuclear Plants

1.4.1.4 SFR Projects in Korea

Korea will base further developments on the KALIMER-600 design. Theconceptual design for KALIMER-600 was finished in 2006 and the advancedconcept is currently being developed. After testing the passive decay heat removalcircuit, an integral testing loop will be constructed. A draft action plan was pre-pared by the Korean government in 2007, and a standard safety analysis report andfinal safety analysis report will be approved by the Korean government. A dem-onstration reactor will be constructed and it is expected to become operational by2028. A comparison of the KALIMER-600 concept with advanced plant specifi-cations is shown in Table 1.7 [42].

1.4.1.5 SFR Projects in India

An extended description of the development in India can be found in [15]. Indiasfast breeder program starts from the existing water reactors. Plutonium and ura-nium reprocessed from them would be effectively utilized in well proven oxidefuel-based fast breeder reactors, and subsequently, at an appropriate stage, whenall the new necessary technologies have been developed and demonstrated,metallic fuel based FBRs will be introduced. India has also envisaged robustthorium reactor technologies as promising sustainable future energy resource.Studies indicate that once the FBR capacity reaches about 200 GWe, thorium-based fuel can be introduced progressively in the FBRs to initiate the third stage ofthe program, where the U-233 bred in these reactors is to be used in the thorium-based reactors. The proposed road map for the third stage therefore comprisesthorium-based reactor technologies, incorporating the (Th-U-233) cycle. India isone of the leading countries in the world in thorium research and has gained thatexperience through thorium irradiation and the operation of U-233 fuelled researchreactors. A 40 MWth Fast Breeder Test Reactor (FBTR) has been in operation inIndia since 1985 [43, 44]. Ref. [44] provides a description of the reactor andsummarizes the operating history of the reactor. It is a loop type sodium cooledfast reactor located at Indira Gandhi center for atomic research (IGCAR),Kalpakkam. The reactor design is based on the French reactor Rapsodie, withseveral modifications, which include the provision of a steam-water circuit andturbine-generator in place of sodium-air heat exchanger in Rapsodie. Heat gen-erated in the reactor is removed by two primary sodium loops, and transferred tothe corresponding secondary sodium loops. Each secondary sodium loop is pro-vided with two once-through steam generator modules. Steam from the fourmodules is fed to a common steam-water circuit comprising a turbine-generatorand a 100 % dump condenser. The reactor uses a high-plutonium mono-carbide asthe driver fuel. Being a unique fuel of its kind without any irradiation data, it wasdecided to use the reactor itself as the test bed for this driver fuel. The FBTR wassynchronized with the grid in July 1997. The operating experience of this FBTR

1.4 Generation IV Nuclear Plants 27

has provided sufficient feedback and confidence for India to launch upon theconstruction of a 500 MWe fast reactor prototype fast breeder reactor (PFBR). ThePFBR, designed by IGCAR, is a 500 MWe, sodium cooled, pool type, mixed-oxide (MOX) fuelled reactor having two secondary loops. Ref. [45] describes thesalient design features including the design of the reactor core, reactor assembly,main heat transport systems, component handling, steam water system, electricalpower systems, instrumentation and control, plant layout, safety, research anddevelopment. The primary objective of the PFBR is to demonstrate techno-eco-nomic viability of FBRs on an industrial scale. The reactor power is chosen toenable adoption of a standard turbine as used in fossil power stations, to have astandardized design of reactor components resulting in further reduction of capitalcost and construction time in future and compatibility with regional grids.

1.4.1.6 SFR Projects in Europe

In Europe, particularly in France, strong interest in SFRs already exists. Withrespect to industrial application, the Superphenix was the most important plant.Also in Germany, a fast sodium breeder reactor project existed (SNR 300 inTable 1.5). However, this plant never went into operation. Currently, several fastreactor concepts are being considered in Europe taking the SFR as the reference

Table 1.7 Concepts for SFRs in South Korea [42] (TBD…to be determined, BOP…balance ofplant, RHRS…reactor heat removal system, PDRC…passive decay heat removal system, SG…steam generator, FMS...ferritic-martensitic steel, TRU.. transuranic)

KALIMER-600

Candidate concepts Advanced concept

Reactor Power, MWe 600 600/900/1200 TBDConversion ratio 1.0 05–0.8, 1.0 05–0.8, 1.0Core exit T., �C 545 510–550 TBDCladding

materialMod. HT9 Mod. HT9/FMS TBD

Fuel type U–TRU–Zr U–TRU–Zr U–TRU–ZrNo. of loops 2 2, 3 TBDReactor vessel

diameter, m11.4 Minimization TBD

In- vesselrotating plug

2 rotatingplugs

2 rotating plugs w/multiwave-guide tubes

2 rotating plugs w/multiwave-guide tubes

SG Tube type Helicalsingletube

Helical single tube/double wall tube

TBD

RHRS PDRC PDRC PDRCSeismic isolation Horizontal Horizontal Horizontal

BOP Energyconversionsystem

Rankine Rankin/S-CO2 Brayton TBD

28 1 Nuclear Plants

technology [46]. In France, the SFR is the candidate prototype of a Generation IVsystem to be built as early as 2020. This project is called ASTRID and mixed oxidefuel (U, Pu) O2 is considered the reference fuel for the core of this reactor [47].The core design of French advanced sodium-cooled fast reactors is mainly drivenby safety, competitiveness, and flexibility margins compared to previous SFRprojects. Performance objectives include improvement of safety features, flexiblemanagement of plutonium (optimization of uranium resources) and transmutationof minor actinides (environmental burden decrease), high burn-up rate, highoperating availability, and proliferation resistance enhancement with integratedfuel cycle. The ASTRID prototype is called a ‘‘self-generating’’ fast reactor ratherthan a breeder in order to demonstrate low net plutonium production. The ASTRIDprogram includes development of the reactor itself and associated fuel cyclefacilities: a dedicated MOX fuel fabrication line and a pilot reprocessing plant forused ASTRID fuel.

1.4.2 Lead-Cooled Fast Reactor

1.4.2.1 Technology Base for the LFR

LFR systems are Pb or Pb–Bi alloy-cooled reactors with a fast-neutron spectrumand closed fuel cycle. One possible LFR system is shown in Fig. 1.14. Optionsinclude a range of plant ratings, ranging from 50 to 150 MWe SMR, and a modularsystem from 300 to 400 MWe. The experience with reactors having lead or leadbismuth as coolants is by far less established than with SFRs. Russia has exper-imented with several lead-cooled reactor designs and has used lead–bismuthcooling for 40 years in reactors for its Alfa class submarines. Existing ferriticstainless steel and metal alloy fuel, which are already significantly developed forsodium fast reactors, are adaptable to Pb–Bi cooled reactors at reactor outlettemperatures of 550 �C. A significant new Russian design is the BREST fastneutron reactor, of 300 MWe or more with lead as the primary coolant, at 540 �C,and supercritical steam generators. A pilot unit is planned at Beloyarsk and1,200 MWe units are proposed. A smaller and newer Russian design is the Lead–Bismuth Fast Reactor (SVBR) of 75–100 MWe. This is an integral design, withthe steam generators sitting in the same Pb–Bi pool at 400–495 �C as the reactorcore, which could use a wide variety of fuels. Temperatures of up to 550 �C ascurrently envisaged are considered as near-term options mainly for electricityproduction. Such plants rely on more easily developed fuel, clad, and coolantcombinations and their associated fuel recycle and refabrication technologies [15,39, 48]. The favorable properties of Pb coolant and nitride fuel, combined withhigh temperature structural materials, can extend the reactor coolant outlet tem-perature into the 750–800 �C range in the long term, which is potentially suitablefor hydrogen production and other process heat applications. In this option, the Bialloying agent is eliminated, and the less corrosive properties of Pb help to enable

1.4 Generation IV Nuclear Plants 29

the use of new high-temperature materials. The required R&D is more extensivethan that required for the 550 �C options because the higher reactor outlet tem-perature requires new structural materials and nitride fuel development. A sum-mary of the design parameters for the LFR systems is given in the following table.Innovations in energy conversion are afforded by rising to higher temperaturesthan liquid sodium (Table 1.8).

This allows going beyond the traditional superheated Rankine steam cycle tosupercritical Brayton or Rankine cycles or process heat applications such ashydrogen production and desalination. The favorable neutronics of Pb and Pb–Bicoolants in the battery option enable low power density, natural circulation-cooledreactors with fissile selfsufficient core designs that hold their reactivities over theirvery long 15–20 year refueling interval. For modular and large units more con-ventional higher power density, forced circulation, and shorter refueling intervalsare used, but these units benefit from the improved heat transport and energyconversion technology. Plants with increased inherent safety and a closed fuelcycle can be achieved in the near- to mid-term.

The longer-term option is intended for hydrogen production while still retainingthe inherent safety features and controllability advantages of a heat transportcircuit with large thermal inertia and a coolant that remains at ambient pressure.

Fig. 1.14 Schematic of a lead cooled fast reactor, Source: US-DOE, http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf

30 1 Nuclear Plants

The favorable sustainability features of fast spectrum reactors with closed fuelcycles are also retained in all options.

1.4.2.2 Materials R&D

The top priority viability R&D areas for higher-temperature starts with materials,screening for cladding, reactor internals, and heat exchangers. The primaryapproach will be to adapt modern materials developments such as composites,coatings, ceramics, and high-temperature alloys from other fields such as aero-space, and gas turbines as stated already in the roadmap [20]. The goal is not onlylong service life but also cost effective fabrication using modern forming andjoining technologies. For the cladding, compatibility with Pb or Pb–Bi on thecoolant side and mixed nitride fuel on the fuel side is required, and radiationdamage resistance in a fast neutron environment is required for a 15–20 yearirradiation period. SiC or ZrN composites or coatings and refractory alloys arepotential options for 800 �C service, while standard ferritic steel is adequate at550 �C. For process heat applications, an intermediate heat transport loop isneeded to isolate the reactor from the energy converter for both safety assuranceand product purity. Heat exchanger materials screening is needed for potentialintermediate loop fluids, including molten salts, He, CO2, and steam. For inter-facing with thermochemical water cracking, the chemical plant fluid is HBr plussteam at 750 �C and low pressure. For interfacing with turbomachinery, theworking fluid options are supercritical CO2 or superheated or supercritical steam.The material screening R&D will take the majority of the viability R&D time

Table 1.8 Different options for liquid metal reactors considered within GIF (see also [20])

Reactor parameters Reference values

Pb-Bi Battery(nearer-term)

Pb–Bi Module(nearer-term)

Pb Large(nearer-term)

Pb (far-term)

Coolant Pb–Bi Pb–Bi Pb PbOutlet temperature

( �C)*550 *550 *550 750–800

Pressure(Atmospheres)

1 1 1 I

Rating (MWth) 125–400 *1,000 3,600 400Fuel Metal Alloy or

NitrideMetal alloy Nitride Nitride

Cladding Ferritic Ferritic Ferritic Ceramic coatings orrefractory alloys

Average burnup(GWD/MTHM)

*100 *100–150 100–150 100

Conversion ratio 1.0 [ = 1 1.0–1.02 1.0Lattice Open Open Mixed OpenPrimary flow Natural Forced Forced NaturalPin linear heat rate Derated Nominal Nominal Derated

1.4 Generation IV Nuclear Plants 31

period and will require corrosion loops, posttest examination equipment, proper-ties testing apparatus, phase diagram development, coolant chemistry controlR&D, fabricability evaluations, and static and flowing in situ irradiation testing.Although these requirements were formulated already 2002 they are still fullyvalid today.

1.4.3 Very-High-Temperature Reactor System

1.4.3.1 VHTR Description

The very-high-temperature reactor system (VHTR) is a next step in the evolu-tionary development of high-temperature gas-cooled reactors. It is a graphite-moderated, helium-cooled reactor with thermal neutron spectrum. The VHTR canproduce hydrogen from heat and water by using thermochemical iodine–sulfur(I–S) process or from heat, water, and natural gas by applying the steam reformertechnology to core outlet temperatures greater than about 950 �C (see Fig. 1.15].It can also make use of electricity and heat for hydrogen production by hightemperature electrolysis. A reference VHTR system that produces hydrogen isshown below. A 600 MWth VHTR dedicated to hydrogen production can yieldover 2 million normal cubic meters per day. The VHTR can also generate elec-tricity with high efficiency, over 50 % at 1,000 �C. Co-generation of heat andpower makes the VHTR an attractive heat source for large industrial complexes.The VHTR can be deployed in refineries and petrochemical industries to substitutelarge amounts of process heat at different temperatures, including hydrogen gen-eration for upgrading heavy and sour crude oil. Core outlet temperatures higherthan 950 �C would enable nuclear heat application to such processes as steel,aluminum oxide, and aluminum production.

The reactor core type of the VHTR can be a prismatic block core such as theoperating Japanese HTTR [49], or a pebble-bed core such as the Chinese HTR-10[50]. For electricity generation, the helium gas turbine system can be directly set inthe primary coolant loop, which is called a direct cycle. For nuclear heat appli-cations such as process heat for refineries, petrochemistry, metallurgy, andhydrogen production, the heat application process is generally coupled with thereactor through an intermediate heat exchanger (IHX), which is called an indirectcycle.

The pebble bed design is based on a fundamental fuel element, called a pebble,that is a graphite sphere (6 cm in diameter- size of a tennis ball) containing a largenumber of uranium oxide particles with the diameter of 1 mm (Fig. 1.16). Theuranium oxide kernel is surrounded by several layers of ceramic coatings. Thestrongest layer is a tough silicon carbide ceramic. This layer serves as a ‘‘pressurevessel’’ to retain the products of nuclear fission during reactor operation or acci-dental temperature excursions. About 330,000 of these spherical fuel pebbles areplaced into graphite core built from graphite blocks. The graphite core is

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constructed from graphite blocks forming an open cylindrical volume. A centergraphite column is placed at the center of the void forming an annular core for thepebbles. The graphite acts:

• as a structure forming the core• as a neutron moderator and reflector, and• as a solid heat absorber and conduction path to ultimate heat sink in case of an

accident.

The graphite core is restrained by lateral restraint straps to keep the graphiteblocks compressed in a cylindrical structure. On the outside of the graphite core isa metallic core barrel that restrains the core during an earthquake and acts as athermal shield to the reactor vessel. The core barrel and graphite core are locatedin a large pressure vessel. Helium gas enters the vessel and flows up in the outerrisers in the permanent graphite reflector reaching the plenum above the corewhere the gas is forced down through the pebbles and out the vessel to thesecondary side of the plant. A small portion the gas in the top plenum flows downthe openings for the control rods cooling them during operation. Another portionof the gas flows down the center graphite column removing heat from there. Thehelium coolant is an inert noble gas that neither severely reacts with materials inthe core at high temperatures nor changes phase with temperature increase.Further, because the pebbles and reactor core are made of refractory materials,

Fig. 1.15 Schematic of a gas cooled high temperature reactor http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf

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they cannot melt and will degrade only at the extremely high temperaturesencountered in accidents (more than 1,600 �C), a characteristic that affords aconsiderable margin of operating safety. The graphite core structure represents alarge thermal capacitance combined with the low power density results in slowthermal transients. Because the pebbles form a packed bed the helium is distrib-uted evenly through without the need of flow channeling. To refuel the pebblecore, pebbles pass through the bottom of the graphite core and new pebbles areadded at the top of the core. This operation is performed continuously duringreactor operations allowing the reactor to stay on line. During operation, onepebble is removed from the bottom of the core about once a minute as areplacement is placed on top. In this way, all the pebbles gradually move downthrough the core like gumballs in a dispensing machine, taking about six months todo so. This feature maintains the optimum amount of fuel for operation withoutrequiring excess activity. It eliminates an entire class of excess-reactivity accidentsthat can occur in current water-cooled reactors. Each expended pebble is measuredto determine the remaining fuel and is stored. The stored pebbles are recycledthrough the core until the remaining nuclear fuel is below a minimum quantity.Also, the steady movement of pebbles through regions of high and low powerproduction means that each experiences less extreme operating conditions onaverage than do fixed fuel configurations, again adding to the unit’s safety margin.After use, the spent pebbles must be placed in long-term storage repositories, thesame way that used-up fuel rods are handled today. The secondary side can supplyprocess heat in the form of steam or another high temperature working fluid.Electricity can be generated directly using a Brayton cycle or with an intermediateheat exchanger using a standard Rankine cycle. Both options can be used simul-taneously where the exhausted heat from Brayton cycle is used as bottomingRankine cycle.

The basic fuel element in a prismatic high temperature gas reactor is a ceramicfuel particle approximately 1 mm in diameter. The spherical fuel particle is a

Fig. 1.16 Pebble type fuelfor a gas cooled hightemperature reactor(� European NuclearSociety, 2003) http://www.euronuclear.org/info/encyclopedia/p/pebble.htm

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ceramic pressure vessel containing a uranium oxy-carbide kernel. The pressurevessel retains the products of nuclear fission during operation or accidental tem-perature excursions. The particles are placed in a fuel compact typically containing4,000–7,000 particles. The fuel compacts are typically 12.7 mm in diameter by50 mm in length. The fuel compacts are pressed into channels drilled into graphiteblocks. There are 14–15 compacts in each channel. Graphite fuel blocks have 210channels; thus, each fuel block contains approximately 3,126 compacts (seeFig. 1.17).

The reactor core consists of an assembly of hexagonal prismatic graphite blocksin annular configuration consisting of three annular rings. The center and outerportions of the core are made from unfueld graphite reflector blocks. The centerring contains the active ring of graphite fuel element blocks. The outer reflectorblocks have full core height channels for control rods. Some of the fuel blocks alsocontain full height vertical channels for control rods and the reserve shut downsystem. The reserve shut down system uses ceramic-coated boron carbide pelletsemploying gravity to fill the channels upon activation. Inherent in the design ofhigh temperature gas reactors is the ability to shut down the reactor during anaccident. As the core heats during an accident, the inherent large negative tem-perature coefficient stops the chain reaction in the active core effectively shuttingdown the reactor. The active core is 10 blocks high with 102 fuel columns. Withthe inner and outer reflector blocks, the physical graphite reactor structure is 6.8 min diameter and 13.6 m high. Graphite pedestals or columns support each graphitecolumn. The area between the columns is the lower plenum. A metallic core barrelrestrains the graphite structure during seismic events and acts as a thermal heat

Fig. 1.17 TRISO coated fuel particles in a prismatic HTR design [51]

1.4 Generation IV Nuclear Plants 35

shield for the reactor vessel. The graphite reactor structure is the solid neutronmoderator and reflector in the core. Graphite remains solid at temperatures wellabove those experienced during accidents. The graphite has a high heat capacitycreating a large heat sink for the core in case of the accident. Further, the high heatcapacity and low power density of the reactor core results in very slow andpredictable temperature transients. The reactor vessel contains the reactor corestructure and shutdown cooling system used for refueling. Helium coolant entersnear the bottom of reactor vessel and flows up the outside of the core barrel to theplenum above the graphite core structure. Helium flows out of the plenum downthrough the coolant holes in the fuel blocks to the lower plenum and out the vessel.The outer and inner reflector has no helium flow with all convection coolingoccurring in the active core. A considerable margin of safety is gained by the useof the inert noble gas helium as the coolant. The gas does not react with the reactorcore materials at high temperatures encountered in accidents (more than 1,600 �C).The helium coolant does not moderate neutron; its use does not add or subtractreactivity.

Refueling the core is handled remotely using a refueling machine located abovethe reactor vessel. A lever arm is attached to an extendable shaft lowered throughan opening in the reactor vessel into the core. The grapple on the end of the leverarm interfaces with the graphite block. Each block is then transferred to a liftstation (another extendable shaft into reactor vessel) where it is pulled up into theshielded refueling machine. The shielded refueling machine then takes the block toadjacent dry storage. The remaining fuel blocks are distributed in the core tocontrol power peaking and flux profile in the core. The fuel cycle is a oncethrough, three-year cycle with one-half of the active core refueled every20 months.

The helium coolant leaving the reactor vessel can be used for process heat aswell as electricity generation. A direct Brayton cycle can use the reactor coolant ina high temperature gas turbine. An indirect Rankine cycle requires an intermediateheat exchanger to transfer heat from the helium coolant to produce steam. Typicalefficiencies of the two cycles depend on the outlet temperature of the reactor. At a700 �C reactor outlet temperature, the Rankine cycle can achieve approximately40 % efficiency. At a higher reactor outlet temperature of 900 �C, the Braytoncycle efficiency is approximately 47 %. Transferring heat for industrial applica-tions requires unique and custom design of heat exchangers to interface with theindustrial application.

The VHTR evolves from HTGR experience and extensive international dat-abases that can support its development (Table 1.9). The basic technology for theVHTR has been well established in former HTGR plants, such as Dragon, PeachBottom, AVR, THTR, and Fort St Vrain and was advanced in concepts such as theGT-MHR and PBMR. The ongoing 30 MWth HTTR project in Japan is intendedto demonstrate the feasibility of reaching outlet temperatures up to 950 �C coupledto a heat utilization process, and the HTR-10 in China will demonstrate electricityand co-generation at a power level of 10 MWth. The former projects in Germanyand Japan provide data relevant to VHTR development. Steam reforming is the

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current hydrogen production technology. The coupling of this technology will bedemonstrated in large scale in the HTTR program but still needs complementaryR&D for market introduction. R&D on thermochemical I–S process is presentlyproceeding in the laboratory-scale stage. Similar to the SFR also for the (V)HTRcurrently demonstration plants exist or are planned which should be briefly dis-cussed in the following.

1.4.3.2 Japan

A demonstrator for a prismatic core is in operation in Japan (HTTR) [49]. Thissystem was originally designed as a heat source for hydrogen production with thethermochemical iodine–sulphur process which was invented by General Atomicsin the 1970s. The main parameters of the plant are summarized in Table 1.10.

On March 13, 2010, long-term (50 days) full power operation of HTTR atreactor outlet coolant temperature of about 950 �C was successfully completed,and various performance data could be obtained. Main future demonstrationactivities will go towards industrialization of the I–S hydrogen process and aHTGR cascade energy plant for 79 % efficient production of hydrogen, electricityand freshwater. A nuclear commercial hydrogen production plant is envisaged by2030.

1.4.3.3 China

China built a pebble bed type of demonstrator (HTR-10) which is based on theformer German experience. The HTR-10 experience shall be used for a new HTR-PM demonstration plant [52]. The HTR-PM plant will consist of two nuclear steam

Table 1.9 Past, current and planned gas cooled reactor projects

U.S. Europe Africa Asia

Past Peach Bottom(P), St. Vrain(P)

AVR (PB), THTR-300 (PB),Germany

Cancelled PBMR (PB)/SouthAfrica

Operating HTR-10 (PB)/ChinaHTTR (P)/(Japan)

Underconstruction

HTR-PN (PB)/China

Planned NGNP

(P...prismatic, PB...pebble)

1.4 Generation IV Nuclear Plants 37

supply systems. Each of these modules consists of a single zone 250 MWthpebble-bed modular reactor and a steam generator. The two modules feed onesteam turbine and generate an electric power of 210 MW. A pilot fuel productionline will be built to fabricate 300,000 pebble fuel elements per year. This line isclosely based on the technology of the HTR-10 fuel.

The main performance data of the HTR-PM are listed in Table 1.11.

1.4.3.4 United States

Very interesting is the development in the US where NGNP should become ademonstrator for electricity and heat generation [53].

Research and development (R&D) specific to NGNP mentioned in the EnergyPolicy Act (2005) and conducted to date is based on the gas-cooled very hightemperature reactor (VHTR) concept promulgated in the Generation IV technol-ogy roadmap [20]. The Very-High-Temperature Reactor (VHTR) system uses athermal neutron spectrum and a once-through uranium cycle. The VHTR system isprimarily aimed at relatively faster deployment of a system for high temperatureprocess heat applications, such as coal gasification and thermochemical hydrogenproduction, with superior efficiency. The reference reactor concept has a600 MWth helium cooled core based on either the prismatic block fuel of the gasturbine–modular helium reactor (GT-MHR) or the pebble fuel of the pebble bed

Table 1.10 Characteristics of the Japanese HTTR

Thermal power 30 MWFuel Coated fuel particle/Prismatic block typeCore material GraphiteCoolant HeliumInlet temperature 395 �COutlet temperature 950 �C (Max.)Pressure 4 MPa

Table 1.11 Performance data of the Chinese HTR-PM [52]

Reactor module numbers 2Thermal power/module 2,250 MWLifetime 40aCore diameter/height 3.0/11 mPrimary system pressure 7.0 MPaHelium inlet/outlet temperature 250/750 �CHelium mass flow 96 kg/sFreshs team temperature/pressure 566 �C/13.2 MPaElectricpower 210 MW

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modular reactor (PBMR). The primary circuit is connected to a steam reformer/steam generator to deliver process heat. The VHTR system has coolant outlettemperatures above 900 �C. It is intended to be a high-efficiency system that cansupply process heat to a broad spectrum of high temperature and energy-intensive,nonelectric processes. The system may incorporate electricity generation equip-ment to meet cogeneration needs.

About 40 % of the US greenhouse gas emissions come from industrialprocesses in high energy consuming sectors. With NGNP systems, the process heator steam generated by the high temperature nuclear reactors will be used to powerapplications such as power generation using advanced highly efficient turbines;plastics manufacturing; petroleum refining and fuels production; and producingammonia for fertilizer. By integrating energy generation and production opera-tions, NGNP technologies will allow high energy consuming industries and sectorsto reduce carbon dioxide emissions, limit their need for fossil fuels, and becomemore competitive. The basic technology for the NGNP has been established informer high-temperature gas-cooled reactor plants shown in Table 1.9.

1.4.3.5 South Africa

The South African PBMR started in 1999 with the development of a direct Braytoncycle plant for electricity generation and low temperature cogeneration applica-tions such as desalination. The plan was to build a demonstration plant called theDPP400 at Eskom’s Koeberg site and the RSA national utility Eskom was thetargetted customer. This plant was designed to generate 165 MW electricity usinga 400 MWt annular core pebble bed reactor coupled to a direct Brayton cyclepower conversion unit.

During the last few years growing interest in HTRs for high temperature pro-cess heat or cogeneration applications became visible. Particularly the US NGNPcould become the first customer for a plant of this type. As a result of thesedevelopments and also of national funding problems the board of PBMR decidedto change to an indirect steam plant which could be used for electricity generationand/or process heat. The current plant design is based on a 2 9 250 MWt reactorlayout where each reactor has its own primary cooling circuit and steam generator.On the secondary side the steam generators are connected to a common steamheader. Although the project advanced quite far, the South African government, inSeptember of 2010, decided to stop funding the effort.

1.4.3.6 South Korea

The South Korean NHDD-project intends to build a VHTR for hydrogen pro-duction. No decision has been taken with respect to core design (block or pebble).The gas outlet temperature is expected to be 950 �C and the reactor power shouldbe 200 MWth. A cold vessel option is considered. Hydrogen shall be produced in a

1.4 Generation IV Nuclear Plants 39

5 train sulfur-iodine thermo-chemical plant. Technology selection should be fin-ished by 2012 and starting of operation of the demonstrator is scheduled for 2026.

1.4.3.7 Materials R&D

Carbon–Carbon Composite Components: Development of carbon–carboncomposites is needed for control rod sheaths, especially for the VHTR based on aprismatic block core, so that the control rods can be inserted to the high-temperatureareas entirely down to the core. Promising ceramics such as fiber-reinforcedceramics, sintered alpha silicon-carbide, oxide-composite ceramics, and othercompound materials are also being developed for other industrial applicationsneeding high strength, high-temperature materials. Necessary R&D includes testingof mechanical and thermal properties, fracture behavior, and oxidation; post irra-diation heat-up tests; and development of models of material behavior and stressanalysis code cases considering anisotropy.

To realize the goal of core outlet temperatures upto 1000 �C, new metallicalloys for reactor pressure vessels have to be established. At these core-outlettemperatures, the reactor pressure vessel temperature will exceed 450 �C. LWRpressure vessels were developed for 300 �C service, and the HTTR vessel for400 �C. Hastelloy-XR metallic materials are used for intermediate heat exchangerand high temperature gas ducts in the HTTR at core-outlet temperatures up toabout 950 �C, but further development of Ni–Cr–W superalloys and other prom-ising metallic alloys will be required for the VHTR. The irradiation behavior ofthese superalloys at the service conditions expected in the VHTR will need to becharacterized. Such work is expected to take 8–12 years and can be performed atfacilities available worldwide. An alternate pressure vessel allowing for largerdiameters and ease of transportation, construction, and dismantling would be theprestressed cast-iron vessel, which can also prevent a sudden burst due to sepa-ration of mechanical strength and leak tightness. The vessel could also include apassive decay heat removal system with enhanced efficiency.

Heat Utilization Systems Materials: Internal core structures and coolingsystems, such as intermediate heat exchanger, hot gas duct, process components,and isolation valve that are in contact with the hot helium can use the currentmetallic materials up to about 1,000 �C core-outlet temperature. For core-outlettemperatures exceeding 1,000 �C, ceramic materials must be developed. Pipingand component insulation also requires design and materials development.

Core Internals: Core internal structures containing the fuel elements such aspebbles or blocks are made of high-quality graphite. The performance of high-quality graphite for core internals has been demonstrated in gas cooled pilot anddemonstration plants, but recent improvements in the manufacturing process ofindustrial graphite have shown improved oxidation resistance and better structuralstrength. Irradiation tests are needed to qualify components using advancedgraphite or composites to the fast fluence limits of the VHTR.

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1.4.4 Gas-Cooled Fast Reactor System R&D

1.4.4.1 GFR Description

The GFR system features a fast-spectrum helium-cooled reactor (Fig. 1.18) andclosed fuel cycle. Like thermal-spectrum helium-cooled reactors such as theGT-MHR and the PBMR, the high outlet temperature of the helium coolant makesit possible to deliver electricity, hydrogen, or process heat with high conversionefficiency. The GFR uses a direct-cycle helium turbine for electricity and can useprocess heat for thermochemical production of hydrogen. Through the combina-tion of a fast-neutron spectrum and full recycle of actinides, GFRs minimize theproduction of long-lived radioactive waste isotopes. The GFR’s fast spectrum also

Fig. 1.18 Schematic of a gas cooled fast reactor (Source: US-DOE) http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf)

1.4 Generation IV Nuclear Plants 41

makes it possible to utilize available fissile and fertile materials (includingdepleted uranium from enrichment plants) two orders of magnitude more effi-ciently than thermal spectrum gas reactors with once-through fuel cycles. TheGFR reference assumes an integrated, on-site spent fuel treatment and refabrica-tion plant.

1.4.4.2 Technology Gaps for the GFR

Although the GFR is in principle based on experience with moderated gas cooledreactors, demonstration of the viability of the GFR requires the solution of anumber of significant technical challenges. Fuel, fuel cycle processes, and safetysystems pose the major technology gaps like:

• GFR fuel forms for the fast-neutron spectrum• GFR core design, achieving a fast-neutron spectrum for effective conversion

with no fertile blankets• GFR safety, including decay heat removal systems that address the significantly

higher power density (in the range of 100 MWth/m3) and the reduction of thethermal inertia provided by graphite in the modular thermal reactor designs

• GFR fuel cycle technology, including simple and compact spent-fuel treatmentand refabrication for recycling.

Performance issues for GFR include:

• Development of materials with superior resistance to fast-neutron fluence undervery-high-temperature conditions

• Development of a high-performance helium turbine for efficient generation ofelectricity

• Development of efficient coupling technologies for process heat applicationsand the GFR’s high temperature nuclear heat.A summary of design parameters for the GFR system is given in Table 1.12

1.4.4.3 GFR Materials R&D

Candidate Materials. The main challenges are in vessel structural materials, bothin-core and out-of-core, that will have to withstand fast-neutron damage and hightemperatures, up to 1,600 �C in accident situations. Ceramic materials are there-fore the reference option for in-core materials, and composite cermet structures orinter-metallic compounds will be considered as a backup. For out-of-core struc-tures, metal alloys will be the reference option. The most promising ceramicmaterials for core structures are carbides (preferred options are SiC, ZrC, TiC,NbC), nitrides (Zr N, TiN), and oxides (MgO, Zr(Y)O2). Intermetallic compounds

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like Zr3 Si2 are promising candidates as fast-neutron reflector materials. Limitedwork on Zr, V or Cr as the metallic part of the backup cermet option should also beundertaken. For other internal core structures, mainly the upper and lower struc-tures, shielding, the core barrel and grid plate, the gas duct shell, and the hot gasduct, the candidate materials are coated or uncoated ferritic-martensitic steels(or austenitic as alternative solution), other Fe–Ni- Cr-base alloys (Incoloy 800),and Ni-base alloys. The main candidate materials for pressure vessels (reactor,energy conversion system) and cross vessel are 2 1/4 and 9–12 Cr martensiticsteels. The recommended R&D activities include a screening phase with materialirradiation and characterization, a selection of a reference set of materials for corestructural materials, and then optimization and qualification under irradiation. Theprogram goal is to select the materials that offer the best compromise regarding:

• Fabricability and welding capability• Physical, neutronic, thermal, tensile, creep, fatigue, and toughness properties

and their degradation under low-to-moderate neutron flux and dose• Microstructure and phase stability under irradiation• Irradiation creep, in-pile creep, and swelling properties• Initial and in-pile compatibility with He (and impurities).

Recommended R&D activities on out-of-core structures consists of screening,manufacturing, and characterizing materials for use in the pressure vessel, primarysystem, and components (pipes, blowers, valves, heat exchangers). With respect tomaterials used for the balance of plant, the development program includesscreening, manufacturing, and characterizing heat-resisting alloys or compositematerials for the Brayton turbomachinery (turbine disk and fins), as well as forheat exchangers, including the recuperator of the Brayton cycle. Likewise, in thecase of nonelectricity energy products, materials development is required forthe intermediate heat exchanger that serves to transfer high-temperature heat in thehelium coolant to the process heat applications.

Table 1.12 Operational design parameters for a gas cooled fast reactor (FIMA...fissions perinitial metal atom)

Reactor parameters Reference value

Reactor power 600 MWhthNet plant efficiency (direct cycle helium) 48 %Coolant inlet/outlet temperatureand pressure

490/850 �C90 bar

Average power density 100 MWth/m3

Reference fuel compound UPuC/SiC (70/30 %) with about 20 % Pu contentVolume fraction Fuel/Gas/SiC 50/40/10 %Conversion ratio Self sufficientBurnup, damage 5 % FIMA, 60 dpa

1.4 Generation IV Nuclear Plants 43

1.4.5 Supercritical Water Reactor

1.4.5.1 SCWR Description

SCWRs are high-temperature, high-pressure watercooled reactors that operateabove the thermodynamic critical point of water (374 �C, 22.1 MPa) [30].A typical SCWR system is shown in Fig. 1.19. These systems may have a thermalor fast-neutron spectrum, depending on the core design. SCWRs have unique fea-tures that may offer advantages compared to state-of-the-art LWRs in the following:

• SCWRs offer increases in thermal efficiency relative to current-generationLWRs. The efficiency of a SCWR can approach 44 %, compared to 33–35 %for LWRs.

• A lower-coolant mass flow rate per unit core thermal power results from thehigher enthalpy content of the coolant. This offers a reduction in the size ofthe reactor coolant pumps, piping, and associated equipment, and a reduction inthe pumping power.

A lower-coolant mass inventory results from the once-through coolant path inthe reactor vessel and the lower-coolant density. This opens the possibility ofsmaller containment buildings. No boiling crisis (a serious issue with PWRs) existsdue to the lack of a second phase in the reactor, thereby avoiding discontinuousheat transfer regimes within the core during normal operation. Steam dryers, steamseparators, recirculation pumps, steam generators are eliminated. Therefore, theSCWR can be a simpler plant with fewer major components. The Japanesesupercritical light water reactor (SCLWR) with a thermal spectrum has been thesubject of the most development work in the last 10–15 years and is the basis formuch of the reference design. The SCLWR reactor vessel is similar in design to aPWR vessel (although the primary coolant system is a direct-cycle, BWR-typesystem). High-pressure (25.0 MPa) coolant enters the vessel at 280 �C. The inletflow splits, partly to a downcomer and partly to a plenum at the top of the core toflow down through the core in special water rods. This strategy provides moder-ation in the core. The coolant is heated to about 510 �C and delivered to a powerconversion cycle, which blends LWR and supercritical fossil plant technology;high-, intermediate and low-pressure turbines are employed with two reheatcycles. The overnight capital cost for a 1,700 MWe SCLWR plant may be as lowas $900/kWe (about half that of current ALWR capital costs), considering theeffects of simplification, compactness, and economy of scale. The operating costsmay be 35 % less than current LWRs.

The SCWR can also be designed to operate as a fast reactor. The differencebetween thermal and fast versions is primarily the amount of moderator material inthe SCWR core. The fast spectrum reactors use no additional moderator material,while the thermal spectrum reactors need additional moderator material in thecore. A summary of designs parameters for the SCWR system is given in thefollowing Table 1.13.

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Fig. 1.19 Schematic of a supercritical water reactor, (Source: US-DOE, http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf)

Table 1.13 Typical design parameters of a SCWR [20]

Reactor parameters Reference value

Plant capital cost 900 $/kWUnit power 1,700 MWe

Spectrum ThermalNet efficiency 44 %Coolant inlet/outlet temperature 280/510 �CPressure 25 MPaAverage power density *100 MWth/m2

Reference fuel UO2

Cladding Ferritic/martensitic steel or nickel-alloyStructural materials (incl. advanced

cladding)Materials development necessary

Burnup *45 GWD/MTHM (Gigawattday/metric ton of heavymetal)

Radiation damage 10–30 dpaSafety Similar to advanced LWR

1.4 Generation IV Nuclear Plants 45

Much of the technology base for the SCWR can be found in the existing LWRsand in commercial supercritical-water-cooled fossil-fired power plants. However,there are some relatively immature areas. There have been no prototype SCWRsbuilt and tested. For the reactor primary system, there has been very little in-pileresearch done on potential SCWR materials or designs, although some SCWRin-pile research has been done for defense programs in Russia and the UnitedStates. Limited design analysis has been underway over the last 10–15 years inJapan, Canada, and Russia. For the balance of plant, there has been development ofturbine generators, piping, and other equipment extensively used in supercritical-water-cooled fossil-fired power plants. The SCWR may have some success atadopting portions of this technology base.

1.4.5.2 Technology Gaps for the SCWR

The important SCWR technology gaps are in the areas of [20]:SCWR materials and structures, including:

• Corrosion and stress corrosion cracking (SCC)• Radiolysis and water chemistry• Dimensional and microstructural stability• Strength, embrittlement, and creep resistance

SCWR safety, including power-flow stability during operation and SCWR plantdesign. Important viability issues are found within the first two areas, and per-formance issues are found primarily within the first and third areas.

Corrosion and SCC: The SCWR corrosion and SCC research activities shouldfocus on obtaining the following information:

• Corrosion rates in SCW at temperatures between 280 and 620 �C (the corrosionshould be measured under a wide range of oxygen and hydrogen contents toreflect the extremes in dissolved gasses)

• Composition and structure of the corrosion films as a function of temperatureand dissolved gasses

• The effects of irradiation on corrosion as a function of dose, temperature, andwater chemistry

• SCC as a function of temperature, dissolved gasses, and water chemistry• The effects of irradiation on SCC as a function of dose, temperature, and water

chemistry.

Radiolysis and Water Chemistry: The SCWR water chemistry research programshould focus on obtaining the following information:

• The complete radiolysis mechanism in SCW as a function of temperature andfluid density

• The chemical potential of H2, O2, and various radicals in SCW over a range oftemperatures (280–620 �C)

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• Recombination rates of various radicals, H2, and O2 in SCW over a range oftemperatures (280–620 �C)

• Effect of radiation type: neutrons, gammas, as well as flux on radiolysis yields• Formation and reaction of other species by radiolytic processes• Impurities introduced into the primary system.

Dimensional and Microstructural Stability: The SCWR dimensional andmicrostructural research activities should focus on obtaining the followinginformation:

• Void nucleation and growth, and the effect of He production, on void stabilityand growth, and He bubble nucleation and growth as a function of dose andtemperature

• Development of the dislocation and precipitate microstructure and radiation-induced segregation as a function of dose and temperature

• Knowledge of irradiation growth or irradiation induced distortion as a functionof dose and temperature

• Knowledge of irradiation-induced stress relaxation as a function of tension,stress, material, and dose.

Strength, Embrittlement, and Creep Resistance: The SCWR strength, embrit-tlement, and creep resistance research activities should focus on obtaining thefollowing information:

• Tensile properties as a function of dose and temperature• Creep rates and creep rupture mechanisms as a function of stress, dose, and

temperature• Creep-fatigue as a function of loading frequency, dose, and temperature• Time dependence of plasticity and high-temperature plasticity• Fracture toughness as a function of irradiation temperature and dose• Ductile-to-brittle transition temperature (DBTT) and helium embrittlement as a

function of dose and irradiation temperature• Changes in microstructure and mechanical properties following design basis

accidents.

1.4.6 Molten Salt Reactor

The molten salt reactor originally proposed in the GIF roadmap [20] was a thermalsystem using graphite as moderator. Meanwhile the high versatility of molten saltled to significant changes (Fig. 1.20). Currently, two baseline concepts are con-sidered [30] which have large commonalities in basic R&D areas, particularly forliquid salt technology and materials behaviour (mechanical integrity, corrosion).These are:

1.4 Generation IV Nuclear Plants 47

• The MSFR (molten salt fast reactor) system operated in the thorium fuel cycle.Although its potential has been assessed, specific technological challengesremain and the safety approach has to be established.

• The FHR (fluoride salt cooled high temperature reactor) system, a high tem-perature reactor with better compactness than the VHTR and passive safetypotential for medium to very high unit power ([2,400 MWt).

In addition, opportunities offered by liquid salts for intermediate heat transportin other systems (SFR, LFR, VHTR) are investigated. Liquid salts offer twopotential advantages: smaller equipment size, because of the higher volumetricheat capacity of the salts, and the absence of chemical exothermal reactionsbetween the reactor, intermediate loop and power cycle coolants. A summary ofcurrently considered concepts is shown in Table 1.14. Liquid salt chemistry playsa major role in the viability demonstration, with such essential R&D issues as: thephysico-chemical behaviour of coolant and fuel salts, including fission productsand tritium; the compatibility of salts with structural materials for fuel and coolantcircuits, as well as fuel processing material development; the on-site fuel pro-cessing; the maintenance, instrumentation and control of liquid salt chemistry(redox, purification, homogeneity), and; safety aspects, including interaction ofliquid salts with various elements.

Fig. 1.20 Molten salt reactor. The GENIV roadmap proposed a thermal spectrum with graphitemoderator (Source: US DOE)

48 1 Nuclear Plants

1.5 Other Advanced Nuclear Plant Concepts

1.5.1 Traveling Wave Reactor

The complexity of the fuel cycle could be eventually considerably reduced when socalled traveling wave reactors could be realized. The TWR is an in situ breederreactor that does not require fuel or blanket reprocessing and recycle [54]. This typeof breeding enables the reactor to operate for decades without refueling, which leadsto very high reactor availability and very low fuel cost over the life of the reactor. TheTWR uses a multi-region core in which a small region containing appropriateamounts of fissile material is made critical to supply excess neutrons to start, or‘‘ignite’’, a breed-burn wave that propagates into adjacent regions containing onlyfertile material. The wave propagates slowly (the order of 1 cm/month) until thewave reaches the end of the fertile regions. A possible realization of such a reactor isshown in Fig. 1.23. The main difference between thermal reactors and fast reactors isthe degree to which uranium can be burned as discussed above. Natural uranium, as itis mined, consists of 0.7 % U235 and 99.3 % U238. Thermal reactors burn primarilyU235, and are able to convert only modest fractions of U238 to Pu239 before theirneutron economies become marginal. As a result, even the best LWRs are able tofission only 0.7 % of all uranium that is mined. Mixed-oxide (MOX) recycling canimprove this use efficiency by about 30 %. In contrast, fast reactors convert U238 tofissile Pu239 or fission U238 directly. Fast reactors can also be designed to createsignificantly more fissile fuel than is used. Because of these abilities, fast reactors areable, in principle, to fission essentially all uranium, as it is mined, provided that thefission products (which parasitically absorb neutrons and thereby progressivelydegrade the reactor’s neutron economy) are removed at least once. Even if fissionproducts are never chemically removed from the reactor, it can be designed to fissionabout 50 % of the natural or depleted uranium before its fuel becomes ‘‘effectivelyspent’’, i.e., no longer capable of producing sufficient neutrons to sustain nuclear

Table 1.14 Fuel and coolant salts for different applications (An represents actinides) [33]

Reactortype Neutronspectrum

Application Carrier salt Fuel system

MSR-Breeder Thermal Fuel 7LiF–BeF27LiF–BeF2–ThF4–UF4

Non-moderated Fuel 7LiF–ThF47LiF–ThF4–UF47LiF–ThF4–PuF3

MSR-Breeder T/NM Secondary coolant NaF–NaBF

MSR-Burner Fast Fuel LiF–NaF LiF–(NaF)–AnF4–AnF3

LiF–(NaF)–BeF2 LiF–(NaF)–BeF2–AnF4–AnF3

LiF–NaF–ThF4

AHTR Thermal Primary coolant 7LiF–BeF2

SFR Intermediatecoolant

NaN03–KNO3–(NaNO2)

1.5 Other Advanced Nuclear Plant Concepts 49

reactions. One example of a fast reactor design that offers such high performancebreeding capability is a TerraPower TWR cooled by liquid sodium (Fig. 1.21). Thisreactor is capable of sustaining energy-producing fission when fueled primarily withnatural or depleted uranium. Only a small amount of enrichment is needed to startfission going, and no chemical reprocessing of spent fuel is ever required. TWRs ofthis kind should be able to achieve a fuel utilization efficiency about 40 times that ofcurrent LWRs. Such a dramatic increase in fuel efficiency has important implicationsfor the sustainability of global uranium resources. Such reactors could operate with awide variety of fuel like depleted uranium or thorium. According to simulations thefollowing commercial applications could be envisaged [55]:

• 1,000 MW net electrical power production• Liquid sodium cooled uranium metal core• Pool-type nuclear island configuration• Compact internal intermediate heat exchanger (sodium-to-sodium)• Rankine steam generator energy conversion• HT-9 fuel pin clad and core support• Boron carbide safety and control rods• Over 30 year core life• Reactor containment based on core damage induced reconfiguration

Fig. 1.21 Travelling wavereactor. Once ignited adeflagration wave will breedfissile material from fertile fueland will burn that material as thewave propagates slowly from oneend of the core tot he other. (withthe permission of IntellectualVentures, see also [54])

50 1 Nuclear Plants

1.5.2 Accelerator-Driven Systems

Energetic particles like neutrons can trigger elemental changes (transmutation).Transmutation could in principle be used to convert long-lived actinides (partic-ularly neptunium, americium and curium) contained in waste from used nuclearfuel into shorter-lived radionuclides. In the concept proposed by Rubbia [56] anaccelerator is combined with a fission reactor to an accelerator driven system(ADS) as shown in Fig. 1.22. Usually protons are accelerated to high energy in acyclotron. The protons hit a target thereby producing spallation neutrons whichcan be used as irradiation source for investigation of radiation damage, fortransmutation of long-lived waste, but also as a power reactor. For this purpose thespallation target can be surrounded by a blanket assembly of nuclear fuel, such asfissile isotopes of uranium. An ADS can be used either for irradiation experimentsor for nuclear reactions in a reactor. High atomic number elements can be usedeither in solid form (tungsten, tantalum etc.) or in liquid form (mercury, lead, lead–bismuth) for targets. A 1,000 MeV beam will create 20–30 spallation neutrons perproton. The target needs to be cooled due to heating caused by the acceleratorbeam. This concept allows to operate the reactor slightly below criticality. Com-pared with conventional reactors the reactor of an ADS can be quickly and reliablycontrolled. The necessary proton beams can be generated with high-current, high-energy accelerators or cyclotrons. An ADS can only run when neutrons are sup-plied to it because it burns material which does not have a high enough fission-to-capture ratio for neutrons to maintain a fission chain reaction.

Fig. 1.22 Accelerator driven system (see also [58])

1.5 Other Advanced Nuclear Plant Concepts 51

The most straightforward and simple ADS-configuration is a liquid metal(Pb, PB–Bi) target/coolant concept in which the primary coolant of the reactor isalso the spallation target. The beam tube and beam window separate the accel-erator beam line vacuum from the target material. The window is positioned at thecentre of the sub-critical core almost at half core height and is cooled by theupward flowing primary coolant under forced convection by the primary pumps.Since the beam window is the most heavily loaded structural part, it must be easilyreplaceable and the possibility of window rupture and its consequences shouldalways be taken into account.

The windowless design would be an alternative. Since any structural material inthe path of the high intensity proton beam will suffer severe radiation damage, a‘‘windowless target’’ design, i.e., without a physical separation between theaccelerator beam line vacuum and the liquid target material is considered.A possible realization is shown in Fig. 1.23 [58]. A nozzle shapes a liquid metal jetso that an optimal target free surface to accept the proton beam is created.A dedicated pump delivering the flow required for the jet formation, is located nearthe edge of the reactor vessel, where more space is available and radiation levels

Fig. 1.23 Principle of awindowless target design [58]

52 1 Nuclear Plants

are low. The heat generated by the proton beam is transported towards the liquidmetal of the primary system by means of a heat exchanger also located near thevessel edge. Nozzle, pump and heat exchanger are combined into a closed loopkeeping most of the spallation products confined and separated from the primarysystem. A windowless design means that the target zone and the beam line share acommon vacuum which leads to several challenging design issues.

In contrast to the liquid metal concept also gas cooled accelerator driven systemsare studied [57]. One concept is the advanced gas cooled accelerator-driven trans-mutation experiment [58]. Such a system offers the following advantages comparedwith liquid metal: less corrosion problems, easy handling of fuel elements, noactivation of the coolant, simplified inspection and repair. Draw backs would be:low heat capacity and high operational pressure (RPV). This concept uses fuel pins.Another concept [59] proposes pebble-type fuel. As far as structural materials areconcerned the ADS needs are comparable with the LMR needs, in case of the liquidmetal option and with the GFR needs, in case of the gas cooled option.

1.5.3 Space Nuclear Plants

A comprehensive summary about nuclear power in space can be found in [55].Using nuclear power for space applications is not new. Applications have beendeveloped in Russia as well in USA since the 1960s. Two systems are considered:

• Radioisotope power sources• Fission power sources

A radioisotope thermoelectric generator (RTG) is a nuclear technology thatuses the heat from radioactive decay. In such a device, the heat released by thedecay of a suitable radioactive material is converted into electricity by the thermo-electric Seebeck effect using an array of thermocouples. Plutonium-238 is used asa heat source because of its high decay heat of 0.56 W/g. RTGs can be consideredas a type of battery and have been used as power sources in satellites, space probesand unmanned remote facilities. RTGs are usually the most desirable power sourcefor robotic or unmaintained situations needing a few hundred watts (or less) ofpower for durations too long for fuel cells, batteries, and in places where solar cellsare not practical. RTGs do not have any moving parts and they are safe, reliableand maintenance-free and can provide heat or electricity for decades under veryharsh conditions, particularly where solar power is not feasible. Figure 1.24 showsa RTG-based general purpose heat source developed by NASA [60].

There were also developments going on for space reactor power systems forheat production and even for space propulsion. Important space reactor powersystems are shown in Table 1.15 which was replotted from [55]. Basically, theseare liquid metal reactors cooled with lithium, sodium or sodium/potassium.

1.5 Other Advanced Nuclear Plant Concepts 53

The energy conversion was either thermo-electric or thermo-ionic (via a hotelectrode which thermoionically emits electrons) [61]. The materials questions,particularly for fuel elements are comparable with non-space applications. Furtherdetails can be found in the literature e.g. [62].

Fig. 1.24 RTG-based general purpose heat Source based for space applications (Source [62])

Table 1.15 Space reactor power systems, t’electric means thermoelectric, t’ionic means ther-moionic (replotted from [57])

Space reactor power systems

SNAP-10 US

SP-100US

RomashkaBouk

BoukRussia

Topaz-1Russia

Topaz-2Russia-US

SAFE-400 US

Dates 1965 1992 1967 1977 1987 1992 2007kWt 45.5 2000 40 \100 150 135 400kWe 0.65 100 0.8 \5 5-10 6 100Converter t’electric t’electric t’electric t’electric t’ionic t’ionic t’electricFuel U-ZrHx UN UC2 U–Mo U02 U02 UNReactor mass,

kg435 5422 45 \390 320 1061 512

Neutronspectrum

Thermal Fast Fast Fast Thermal Thermal/epithermal

Fast

Control Be Be Be Be Be Be BeCoolant NaK Li None NaK NaK NaK NaCore temp. �C,

max585 1377 1900 ? 1600 1900? 1020

54 1 Nuclear Plants

1.5.4 Nuclear Fusion

Nuclear fusion has been considered as a possible sustainable future energy sourcesince quite a while. In contrast to nuclear fission nuclei are fused in a fusion plant.Different candidates for fusion reactions are possible [63, 64]. Currently the fusiontest reactor ITER [65, 66] is built in Cadarache in France. In this device Deuteriumand Tritium are are fused to helium thereby emitting neutrons. The d ? t reactionemployed in current devices emits neutrons with a peak at 14 MeV. The neutronenergy spectrum of the d-t reaction is shown in Fig. 1.25 [67] together with twoother possible fusion reactions. A challenge for fusion concepts concerns the totalenergy balance. In a fusion power reactor a plasma must be maintained at a hightemperature enabling nuclear fusion to occur. The fusion energy gain factor,usually expressed with the symbol Q, is the ratio of fusion power produced in anuclear fusion reactor to the power required to maintain the plasma in steady state.The condition of Q = 1 is referred to as break-even.

The development of plasma systems can be seen from Fig. 1.26 [66] where theplasma temperature is plotted as a function of the Fusion Triple Product which isthe product of density, temperature, and confinement time. It can be seen that thecurrent ITER project is expected to reach ignition conditions. The basic principleof a fusion power plant can be seen in Fig. 1.27 [67]. The D ? T fusion reactiontakes place in the vessel in the center. In contrast to Deuterium Tritium is not

Fig. 1.25 Neutron spectrum Neutron energy spectrum for d ? d, d ? t and t ? t fusionreactions ITER will run a d ? t process (Source [69])

1.5 Other Advanced Nuclear Plant Concepts 55

abundant and therefore for final fusion plants it is planned to breed it from lithiumin a blanket. For the ITER project only tests into this direction are forseen. Theheat is coupled out from the blanket through a heat exchanger or steam generator.The steam or hot gas (e.g. helium) will than drive the turbo-machine for electricitygeneration. It can also be used for thermal processes. High temperatures, highradiation and the need to convert the heat produced into electricity or process heatis rather similar to advanced fission plants. Therefore many similarities in struc-tural materials problems exist. Fusion projects are rather expensive and thereforedifferent schedules for further project development exist. The maximum path fo-reward would be the ‘‘fast track’’. It builds on the ITER plant (currently underconstruction in Cadarache, France) and the irradiation facility IFMIF which wouldallow advanced materials development. Under these assumptions a demonstrator(DEMO-PROTO) could be realized in 30-50 years. The development steps toge-ther with the main parameters of the different plants are shown in Fig. 1.28. In thisconcept a radiation source for the development of radiation resistant and lowactivation materials is included. In principle structural materials for fusion do notdiffer much from structural materials for advanced fission. Main challenge is thehigh exposure to damage from neutrons having an energy spectrum peaked near14 MeV with annual doses in the range of 20 dpa (displacement per atoms), andtotal fluences of about 200 dpa. To minimize nuclear waste these materials(mainly steels) must contain only low activation alloying elements. Testing ofcandidate materials requires a reliable high-flux source of high energy neutronswhich is currently not available. An accelerator-based neutron source has been

Fig. 1.26 Development steps in controlled fusion [68]

56 1 Nuclear Plants

established through a number of international studies and workshops as anessential step for material developing and testing (IFMIF).

The results gained with ITER and IFMIF should provide the basis for the fusiondemonstration plant DEMO/PROTO later. Although at a first glance there seem tobe only limited similarities between nuclear fusion and nuclear fission plants with

Fig. 1.27 Fusion power plant (after [67])

Fig. 1.28 Fast track concept to a fusion demonstration plantin 30–40 years

1.5 Other Advanced Nuclear Plant Concepts 57

respect to structural materials it turns out that the materials needs are comparablealthough fusion plants have to accomodate the very high surface temperaturesoccurring in the plasma facing components.

1.6 Conversion of Nuclear Energy into Electricity and Heat

Current LWRs are premarily used for production of electric energy with a steamturbine which is either part of the primary coolant cycle (BWR) or which isproduced with a steam generator in a second loop (PWR). The high coolanttemperatures, particularly the ones of gas cooled reactors, allow more efficientconversion cycles like a direct cycle helium gas turbine or a supercritical steamcycle. Improvement of efficiency of nuclear power plants and co-generation ofelectricity and heat was one of the goals of the GIF. Particularly the VHTR wasoriginally considered as heat source for thermo-chemical hydrogen productionwith the iodine sulphur (I/S) process. Since the date of publication of the GENIVroadmap a wider range of industrial processes, which could be driven with nuclearheat, has been considered. Figure 1.29 shows different industrial processes and therequired temperatures. Current trends go towards the use of de-centralized small-medium sized reactors which can be considered as energy source at site for dif-ferent industrial processes.

Fig. 1.29 Possibilities for process heat applications

58 1 Nuclear Plants

The main driving force behind these concepts is the reduction of the emission ofgreenhouse gases from industrial processes. Hydrogen is not only an importantenergy carrier for direct use in automotive applications. It is also an importantresource for make-up of fossil resources and other refinery processes. Veryinteresting combined cycle processes using nuclear energy for cogeneration ofelectricity with coal, CO2-capture and Fischer–Tropsch methanol synthesis arealso under consideration. Projects are under discussion where a nuclear driven hightemperature electrolysis splits water into hydrogen and oxygen. The oxygen couldbe used for high temperature coal gasification with CO2 removal. This CO2 couldbe converted with the hydrogen into methanol. Metallurgical reduction is usuallydone with carbon and high amount of CO2 ir thereby created. Hydrogen couldtherefor play an important role in metallurgical processes where this carbonaceousreduction could be replaced by hydrogen reduction. A vision of the use of elec-tricity and heat provided by the US NGNP is shown in Fig. 1.30. The nuclear unitis considered as a part of a CO2-free source for electricity and heat in an advancedchemical plant environment. Electricity and heat are the major contributionscoming from the high temperature reactor.

Process steam is another product which can be supplied from nuclear power.Steam, which is currently mainly generated by fossil fuel, is also used in chemicalplants and for extracting oil from sands/shale. Nuclear heat is also considered asheat source for seawater desalination plants.

Three process routes for production of hydrogen with a nuclear energy sourceare currently considered:

Fig. 1.30 The US NGNP as energy source for industrial applications (Courtesy of IdahoNational Laboratory)

1.6 Conversion of Nuclear Energy into Electricity and Heat 59

• Electrolysis at low temperatures using LWR’s (including waste heat)• High Temperature Electrolysis• Direct thermochemical production (e.g. Iodine/sulphur process)

Several direct thermochemical processes are being developed for producinghydrogen from water. For economic production, high temperatures (800–1,000 �C)are required to ensure rapid throughput and high conversion efficiencies. In severalleading thermochemical processes the endothermic decomposition of sulfuric acidinto oxygen and sulfur dioxide

H2SO4¼¼ [ H2Oþ SO2 þ 1=2O2

plays a central role. Most attractive is the iodine sulfur (IS) process in which iodinecombines with the SO2 and water to produce hydrogen iodide which then disso-ciates to hydrogen and iodine. This is the Bunsen reaction and is exothermic,occurring at low temperature (120 �C):

I2 þ SO2 þ 2H2O ¼¼ [ 2HIþ H2SO4

The HI then dissociates to hydrogen and iodine at about 350 �C, endothermically:

2HI ¼¼ [ H2 þ I2

This can deliver hydrogen at high pressure. The net reaction is then:

H2O ¼¼ [ H2 þ 1=2O2

All the reagents other than water are recycled; there are no effluents. Thisprocess has been successfully demonstrated at laboratory scale. Upscaling toproduction level is currently studied in several countries.

Lacking experience with combination of nuclear and non-nuclear plants withrespect to risk assessments and safety culture is a challenging issue to be con-sidered for licensing of nuclear/non-nuclear combined installations. In this contextit is worthwhile mentioning that nuclear-nonnuclear coupling is also done withcurrent LWRs. District heating is a well known example in this respect. But thereare also other applications using steam from nuclear power plants. For examplehas the Swiss nuclear power station Gösgen been supplying a papermill with steamsince 1979. The mill is located about 1.5 km away from the nuclear power station.The evaporator is heated with steam taken from the nuclear power plant betweensteam generator and turbine. The feedwater from the papermill goes back to theevaporator. Iodine-131 is monitored for safety reasons.

Also combined processes between renewable energy resources and nuclearpower are considered. Nuclear energy with its base-load capabilities could be aninteresting supplement to the inherently cyclic operating renewable energy plantsbased on solar or wind.

60 1 Nuclear Plants

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