New Systems to Curbwaste at Source?

Advances for tomorrow and after

Results obtained in the context of investigations on separation and transmutation make it possible to contemplate the gradual deployment – taking the opportunity provided by a coming renewal of the nuclear power generation fleet – of systems having the capability of fully recycling actinides. With a consequent drastic reduction in the constraints relating to ultimate waste, and enhanced proliferation resistance. New systems to curb waste at source?

cycle”systems, to be optimized as a whole. The other major goal of fourth-generation systems is to open up the field of nuclear power applications, to the pro- CEA/DEN duction of fuels for transportation (hydrogen, and synthetic hydrocarbons), and supply of high-temperature heat to industry. This goal should, in the longer term, also include such advances as may be achieved as regards waste management (see Box 1). In France,while recognizing the validity of these various goals for the nuclear systems of the future, the Atomic An instance of integration of a fourth-generation reactor – in this case, Energy Committee (Comité de l’énergie atomique), a very-high-temperature reactor (VHTR) – into an electricity and hydrogen cogeneration complex. The cylindrical structure positioned next to the reactor is a heat exchanger. in its meeting of 17 March 2005, considered priority The interposition of a hillock between the hydrogen plant and the buried reactor bears should be given to investigations on fast-neutron reac- witness to the precautions used to manage both the nuclear and the hydrogen risks. tors (or fast reactors: FRs) and advanced cycle processes,emphasizing the importance attached to the use of uranium, and the form of ultimate waste, for the A fourth generation of nuclear systems, second stage of renewal of the reactor fleet, around for what purpose? 2040.

Management of high-level radioactive waste stands Which nuclear technologies as a major target for advances, for the nuclear sys- for a sustainable development? tems of the future. Making good use of uranium, and not merely 0.5% of that element (uranium 235), As a whole, advances anticipated for the fuel cycle of as is the case for the reactors of today, stands as the fourth-generation systems concern three areas: raising prime goal, if nuclear power is to be able to make a the energy potential of nuclear fuel reserves,first of all, significant contribution to meeting the world’s energy with the ability to convert,and burn,fertile materials, requirements, and thus help curb greenhouse gas by means of fast neutrons,and recycle the fuel.Optimum emissions. Recycling all fuel to minimize the pre- use of storage and/or disposal sites, second, by mini- sence, in waste, of materials active over very long mizing nuclear waste volume and thermal load.A sim- timespans – minor actinides (MAs) – is the second plification of the demonstration of performance for major challenge. Such challenges, just as the neces- such sites, finally, through a significant reduction in sary advances in terms of economic competitiveness, the lifespan and potential toxicity of radioactive waste, safety, reliability, proliferation resistance, and phy- through the transmutation of long-term contributor sical protection, stand among the goals assigned, by elements, taking advantage of recycles inside reactor 2000, to the reactors of the latter half of the 21st cen- cores. tury, by the Generation IV International Forum.To Fast-neutron nuclear systems are more effective than achieve this, these reactors must use fast neutrons, thermal-neutron systems, such as PWRs, as regards and recycle all of their fuel – the characteristics of use of plutonium, and transmutation of minor acti- four of the six systems selected, at the end of 2002, nides. This advantage is due to the physical characte- by the Forum (sodium-, gas-, lead-, and supercriti- ristics of neutron reactions with transuranic elements. CEA cal-water-cooled fast reactors). The term “system,” Indeed,the probability of inducing fission (which des- Design project used to refer to these, marks the importance atta- troys the nucleus) rather than a neutron capture (which for a fourth-generation gas-cooled fast-neutron ching to the fuel cycle and its optimization, for the yields higher isotopes) is much higher in fast-neutron reactor (GFR). reactors of the future, conceived as “reactor, fuel and systems than in thermal-neutron reactors, by a factor

96 CLEFS CEA - No. 53 - WINTER 2005-2006 Fuel cycle strategies adopted in other countries 1

nuclear- NP 2010 generated Generation IV and Nuclear hydrogen initiative hydrogen

availability of the fast

production reactor nuclear power nuclear deployment demonstrator first option of first U.S. of NGNP** commercial ALWR* NGNP

AFCI opening of Yucca Mountain site fast transuranic burners self-sustaining

management recommendation ultra-high- proliferation-resistant recycling closed high-level waste high-level for a second site burnup fuel Pu + Np (+ Am?) in thermal reactorsof fuel cycle once-through fuel cycleadvanced thermal reactor fuel cycle actinide recycle 2000 2010 2020 2030 2040 2050 * Advanced Light-Water Reactor, of the EPR or AP1000 type. ** Next-Generation Nuclear Plant (gas-cooled thermal reactor)

Figure E1. The United States’ long-term nuclear power development strategy.

In the United States, since 1997, a num- ration power plants (Nuclear Power 2010 a closed fuel cycle, both in order to opti- ber of official committees (PCAST, Program); and the launching, in 2003, of mize waste management, and to improve NERAC…) have been at work on a strategy the AFCI (Advanced Fuel Cycle Initiative) the security of its energy supply (see to provide renewed impetus for nuclear Program, covering reprocessing, recy- FigureE2). The study anticipates, in par- power. These initiatives led DOE to take a cling, and transmutation (see Figure E1). ticular, a demonstration, between 2015 number of major decisions: the setting up, In Japan, JNC published, in 2000, in col- and 2030, of recycling, on a significant scale, in 2000, of the Generation IV International laboration with the Japanese power utili- of actinides (both major and minor) in this Forum; the announcement, in 2002, of a ties, a feasibility study for various fast-neu- type of reactor. Negotiations on the moda- number of measures to reinvigorate cons- tron reactors, showing the need, for that lities of such an international demonstra- truction, in the short term, of third-gene- country, to turn to these technologies, using tion have begun, with active participation from CEA, under the aegis of the Generation IV International Forum. Stages in the development of FR–cycle system Among other initiatives relating to systems of the future, the International Project on 2000 2005 2010 2015 2030 Innovative Nuclear Reactors and Fuel Cycles (INPRO), launched by IAEA in phase I phase II phase III September 2000, is mainly concerned with evaluation selection engineering engineering defining an evaluation methodology for of of promising scale scale various combination tests tests nuclear systems, in particular with respect options of fast to safety, proliferation, the environment, reactors and conceptual detailed and waste management. fuel cycle design design phase IV In Europe, a program on systems of the systems studies studies overall future, organized along the lines of the six demonstration pathways identified by the Generation IV of FR–cycle fast-neutron power reactor deployment International Forum, was initiated in the irradiation reliability 6th Framework Program (FP6); further, a proposal has been submitted for a “Fission” Monju Joyo Platform, under FP7 (2007–11), to cover prototype reactor experimental current research on water-cooled reac- reactor JOYO MONJU tors, and two directions for innovation for future systems: fast neutrons and recycling for sustainable nuclear power, and chemical fabrication recycle high temperature for cogeneration and processing equipement CPFof transuranic RETF heat-generation applications. facility fuel fabrication test Figure E2. testing of engineering scale facility Japan’s long-term nuclear power key technologies verification development strategy.

CLEFS CEA - No. 53 - WINTER 2005-2006 97 98 LF E o 53- WINTER2005-2006 CLEFS CEA-No.

cross-section (barns) 10,000 Advances for tomorrow andafter 1,000 energy. as afunctionofneutron f cross-sections Capture andfission Figure 1. or americium241 0.01 100 0.1 10 10 1 -5 10 -4 .0 .10111 0 1,000 100 10 1 0.1 0.01 0.001 iue3.Ti hnei iecl,combined with Thischange intimescale, Figure 3). co current practice, asisthecasewith than 10,000years, rather over 300–500years, the initialfuelwasdrawn, co W te andto reduce itslong- quite significantly Figure 2), ultimate waste solelyto tially, re actinides.Fully fullrecycling allowing of cling case,thus Cf…)inthemultiple recy- Bk, isotopes (Cm, higher concentration atanequilibrium for naturally arriving in infast-neutron systems, results, This characteristic Boxes EandF). around 10forthefirstinteraction (seeFigure 1)(see of the third generation ofnuclear power generation reactors. Construction site,inFinland,for theOlkiluoto3reactor, thefirst EPRreactor tobebuilt.EPRswillform, inFrance, rm yln h ulmksi osbet etit essen- cycling thefuelmakes itpossibleto restrict, aste of this quality reverts to thisquality apotential harmfulness aste of prbet hto the tomparable thatof nsisting inleaving minoractinidesinthewaste (see harmfulness and harmfulness energy E(eV) r fission capture esidual natural uranium natural 10 fission products thermal thermal 4 10 5 power 10 from which 6 10 . 7 (see used. andasto theprocesses thatshouldbe management, over of the international scene, ataunifiedview,on should alsocontribute to arriving Thisplatform valid forotherfast-neutron systems). demonstrations,equally abroad common(with core of involved infullfuelrecycling insodium-cooled FRs forallprocesses platform international experimental tr involving coun- three partner These demonstrations, PACE the of undertheaegis tion processes (research groups, inJapanin theMonju by reactor,operated w investigations candraw on States ( blies gr kilo- several tens of of by thatlaboratory, separation, tr for thedemons- shouldbesetup, national program tr fuelsincluding of thefabrication of demonstrations to enable shouldbesetupatLaHague, laboratory the outin carried investigations onenhanced separation, from deriving step(in the inFigure 2), “reprocessing” ino the tion of to such processes isto passthrough three majormiles- of thefuelto berecycled.Development and refabrication of calls fornovel processes forspentfuelreprocessing, Fu marked standsasavery advance. thousand years, co ansuranic elements.Around 2020–25,finally,an inter- 2020–25,finally,an elements.Around ansuranic to foealatnd aaeet includingthe overall actinidemanagement, ation of ies (France, Japan, United States), will standasan will United States), Japan, ies (France, ith theFrench e.Tefrtsol e rud21,ademonstra- around 2012, Thefirstshouldbe, nes. m fmnratnds arcto ffuel of fabrication minoractinides, ams of nfinement lrccigo allactinidesinfast-neutron systems ll recycling of A talante nldn rnuai lmns intheUnited elements, including transuranic Pr ORNL ogram). Ganex aoaoy n21–0 aninternational In 2015–20, laboratory. or aigalfsa fseveral hundred having alifespanof CNRS INL gr ouped actinide extraction process extraction actinide ouped rFac,n raso such fuel of trials ) orFrance,and on separation andtransmuta- on separation a collaborative agreement l,optimized actinide all, J AEA assem- .T hese

TVO/Hannu Huovila Towards recycling of waste U in fourth-generation systems? nat

The availability, in France, around 2040, of fast-neutron reactors, together with a spent fuel reprocessing ultimate process allowing the separation of minor actinides, spent fuel reprocessing and waste makes it possible to contemplate deployment of a trans- refabrication FPs mutation strategy that would take advantage of the replacement of industrial installations for nuclear power generation (FRs), and reprocessing (the La Hague Gen IV fast neutrons plant). This approach allows optimization of current actinides facilities, and taking the opportunity of their replacement to switch processes and technologies. A reference scenario (see Figure 4) anticipates,concur- Figure 2. rently, around 2040, initial deployment of fast-neu- Principle schematic of full actinide recycling. tron systems in the French fleet (whether sodium- or gas-cooled; see Box 2), and the coming on stream of a 10,000 new spent fuel reprocessing plant (UP4), supplanting the current one at La Hague.This scenario leaves open MAs 1,000 Pu the options of solely recycling U and Pu, or going for FPs the full recycling of actinides (U–Pu–MAs), depen- MAs ding on the targets set for advances with respect to ulti- plutonium FPs mate waste quality,and overall techno–economic opti- 100 recycling mization of the cycle back-end.Ongoing investigations direct disposal of spent fuel tend to show this strategy is both flexible and robust, 10 in many respects. elative radiotoxicity elative

It leaves the options open, first of all, regarding mana- r natural uranium gement of minor actinides. The existence of conside- 1 MA rable common ground,as regards the various actinide separation and separation and conversion processes,should allow the FPs transmutation new reprocessing plant to be so dimensioned as to 0.1 10 100 1,000 10,000 100,000 1,000,000 continue carrying out sole U–Pu recycling, if requi- time (years) red, while making possible, in the longer term, recycling of all actinides, whether in two separate streams Figure 3. (U–Pu, and MAs), or in a single stream, with joint Evolution of the potential harmfulness of ultimate waste according to various recycling modes. management of all actinides (U–Pu–MAs),if this option proves compatible with the cycle optimization refer- ket, or homogeneous,with U–Pu–MAs in the fuel). red to above. Recycling of minor actinides may be Fast-neutron systems can clear the accumulated acti- contemplated as soon as fast reactors are deployed, nide stock,currently stored either in cooled spent fuel, provided by that time construction is in hand of fabri- or in specific storage facilities. Such a clearance is pos- cation workshops for transuranic element-bearing sible,due to the ability of fast-neutron reactors to accept fuels, corresponding to the technology of the selected 2.5–5% minor actinides in their core. pathway, and designated recycling mode (heteroge- This strategy may further be adjusted to cater for defer- neous,with U and Pu in the fuel,and MAs in the blan- red deployment of FRs. The new reprocessing plant

reactor lifespan Gen IV extension

current fleet EPRs

cycle

U (storage) Gen IV recycling UP4 Figure 4. Pu (MOX monorecycling) Replacement scenario for the French nuclear power MAs & glass or Gen IV generation fleet, and the MAs + FPs & glass disposal FPs & glass La Hague reprocessing plant. The current fleet is comprised of second- generation reactors, 1975 2000 2025 2050 2075 EPRs standing as the third generation.

CLEFS CEA - No. 53 - WINTER 2005-2006 99 Advances for tomorrow and after

“Cool” fuel for high temperatures 2

Feasibility of the gas-cooledfast-neutron reac- ▲ tor(GFR) entails development of a fuel exhibiting particularly high performance, in terms of 500 °C 6 behavior in the presence of fast neutrons and ▲ at high temperature, affording a true field of innovation for design, and materials. In particular, the size of the coolant gas passage sec- 4 (U,Pu)C tion calls for selection, for the fuel, of fertile/fissile materials with a high density of heavy nuclei; while the concern to restrict heating up, and

thereby release of gaseous fission products, at. / (U + Pu) at.) 2 ◆

-3 ◆ 1,250 °C fission gas generation line rising pressure of which stresses the cladding, 1,500 °C

(10 ● leads to considering a compound exhibiting very ● ● ●● 1,750 °C

fission gases present in fuel gases present fission ◆ ◆ high heat conductivity. The high operating tem- ◆ ■ ■ perature (warranting high efficiency, and an 0 6 0 24 opening up to cogeneration), which may even become very high in accident conditions burnup rate (%) (< 1,600 °C), requires the selection of refractory materials, exhibiting reasonable heat conduc- Figure E1. Retention of fission gases by a uranium- and tion, along with the development of suitable plutonium-carbide-based fuel, as a function fabrication and reprocessing processes. of burnup rate. As regards the actinide compound, carbides and nitrides exhibit markedly better heat conductivity than oxides (# 20 W · m– 1 · K– 1 at 1,000 °C, i.e. 3 times higher than for oxides). As regards the other materials for a composite fuel, since the latter is often called upon to ensure a function of fission product confinement (see Figure E1), the high temperature leads to loo- king to ceramic materials, particularly compo- sites featuring a fiber-based structure, for good mechanical resistance. Compounds being inves- tigated include SiC, TiC, TiN, and Ti SiC , the 3 2 Figure E2. (à gauche) latter exhibiting very high ductility. Pin fuel for the gas-cooled fast-neutron reactor (GFR). Fuel development is basically focusing on two design concepts: one being a pin design, featu- Figure E3. (à droite) ring a simple, well-known geometry, making it Plate fuel for GFR. possible to provide for expansion spaces at the extremities, for fission gases (see Figure E2), the other involving honeycomb plates, featuring in the cavity holding it, and, probably, good a large surface area for heat exchanges, confi- control of the mechanical interaction between nement of fission products around each pellet, fuel and confinement structure (see Figure E3).

may be fitted by 2040 with functionalities allowing it contemplate a gradual transmutation strategy. With to cater, as and when required, for the various recy- the ability to adjust to the conditions governing deploy- cling modes possible with later fast-neutron systems, ment of fast-neutron systems in the French fleet, this while not closing the option of switching, at a later would preserve the capability, if this proves compati- date, to full actinide recycling, which would allow a ble with overall techno–economic cycle optimization, reduction by two orders of magnitude of residual quan- for full actinide recycling,affording the ability to achieve tities of actinides in waste. a radical reduction in the potential radiotoxicity and One possible strategy would be to deploy,around 2040, thermal load of ultimate waste,while making the entire an initial series of fast-neutron systems using pluto- fuel cycle more proliferation resistant. nium from the PWR fleet, which might stand as pre- cursors for a fourth generation of reactors,to be deployed > Frank Carré and Philippe Brossard as and when use of breeder reactor technology beco- Nuclear Energy Division mes a requisite. CEA Saclay Center

A gradual transmutation strategy?

The scientific and technical findings obtained in the context of investigations on separation and transmutation, and the opportunities provided by the replacement of nuclear installations mean it is possible to

100 CLEFS CEA - No. 53 - WINTER 2005-2006 A What is radioactive waste?

ccording to the International Atomic a few hundred to one million Bq/g, of traceability), predicting future waste pro- AEnergy Agency (IAEA), radioactive which less than 10,000 Bq/g is from long- duction trends to 2020, and informing the waste may be defined as “any material for lived radionuclides. Its radioactivity beco- public (see An inventory projecting into the which no use is foreseen and that contains mes comparable to natural radioactivity future). ANDRA published this reference radionuclides at concentrations greater in less than three hundred years. National Inventory at the end of 2004. To than the values deemed admissible by the Production of such waste stands at some meet requirements for research in com- competent authority in materials suitable 15,000 m3 per year in France; pliance with the directions set out in the for use not subject to control.” French law – long-lived low-level waste (LLW-LL); French Act of 30 December 1991 (see in turn introduces a further distinction, this category includes radium-bearing Radioactive waste management research: valid for nuclear waste as for any other waste from the extraction of rare earths an ongoing process of advances), ANDRA, waste, between waste and final, or “ulti- from radioactive ore, and graphite waste in collaboration with waste producers, mate,” waste (déchet ultime). Article L. from first-generation reactors; has drawn up a Dimensioning Inventory 541-1 of the French Environmental Code – long-lived intermediate-level waste Model (MID: Modèle d’inventaire de thus specifies that “may be deemed as (ILW-LL), this being highly disparate, whe- dimensionnement), for the purposes of waste any residue from a process of pro- ther in terms of origin or nature, with an arriving at estimates of the volume of duction, transformation or use, any sub- overall stock standing, in France, at waste packages to be taken on board in stance, material, product, or, more gene- 45,000 m3 at the end of 2004. This mainly research along direction 2 (disposal). This rally, any movable property left derelict or comes from spent fuel assemblies (clad- model, including as it does predictions as that its owner intends to leave derelict,” ding hulls and end-caps), or from opera- to overall radioactive waste arisings from further defining as ultimate “waste, be it tion and maintenance of installations; this the current reactor fleet, over their entire the outcome of waste treatment or not, includes, in particular, waste conditioned lifespan, seeks to group waste types into that is not amenable to further treatment during spent fuel reprocessing operations families, homogeneous in terms of cha- under prevailing technological and eco- (as from 2002, this type of waste is com- racteristics, and to formulate the most nomic conditions, in particular by extrac- pacted, amounting to some 200 m3 plausible hypotheses, with respect to tion of the recoverable, usable part, or annually), technological waste from the conditioning modes, to derive the volu- mitigation of its polluting or hazardous operation or routine maintenance of pro- mes to be taken on board for the purpo- character.” duction or fuel-processing plants, from ses of the investigation. Finally, MID sets Internationally, experts from IAEA and the nuclear reactors or from research cen- out to provide detailed stocktaking, inten- Nuclear Energy Agency (NEA) – an OECD ters (some 230 m3 annually), along with ded to cover waste in the broadest pos- organization – as those in the European sludges from effluent treatment (less than sible fashion. MID (not to be confused with Commission find that long-lived waste pro- 100 m3 annually). Most such waste gene- the National Inventory, which has the duced in countries operating a nuclear rates little heat, however some waste of remit to provide a detailed account of power program is stored securely nowa- this type is liable to release gases; actual waste currently present on French days, whilst acknowledging a final solu- – high-level waste (HLW), containing fis- territory) thus makes it possible to bring tion is required, for the long-term mana- sion products and minor actinides parti- down the variety of package families to a gement of such waste. They consider burial tioned during spent fuel reprocessing (see limited number of representative objects, in deep geological structures appears, pre- Box B), and incorporated at high tempe- and to specify the requisite margins of sently, to be the safest way to achieve final rature into a glass matrix. Some 120 m3 error, to ensure the design and assess- disposal of this type of waste. of “nuclear glass” is thus cast every year. ment of disposal safety will be as robust This type of waste bears the major part as feasible, with respect to possible future What constitutes radioactive of radioactivity (over 95%), consequently variations in data. waste? What are the volumes it is the seat of considerable heat release, To ensure consistency between investi- currently involved? this remaining significant on a scale of gations carried out in accordance with Radioactive waste is classified into a num- several centuries. direction 2 and those along direction 3 ber of categories, according to its level of Overall, radioactive waste conditioned in (conditioning and long-term storage), CEA radioactivity, and the radioactive period, France amounts to less than 1 kg per year, adopted MID as input data. MID subsu- or half-life, of the radionuclides it per capita. That kilogram consists, for mes waste packages into standard pac- contains. It is termed long-lived waste over 90%, of LILW-SL type waste, bearing kage types, then computes the number when that period is greater than 30 years, but 5% of total radioactivity; 9% of ILW- and volume of HLW and ILW-LL packa- short-lived waste otherwise. The French LL waste, less than 1% HLW, and virtually ges, according to a number of scenarios, classification system involves the follo- no LLW-LL waste. all based on the assumption that current wing categories: nuclear power plants will be operated for – very-low-level waste (VLLW); this What of the waste of tomorrow? 40 years, their output plateauing at contains very small amounts of radionu- From 1991, ANDRA compiled, on a yearly 400 TWhe per year. clides, of the order of 10–100 Bq/g (bec- basis, a geographical inventory of waste Table 1 shows the numbers and volumes querels per gram), which precludes consi- present on French territory. In 2001, for each standard package type, for the dering it as conventional waste; ANDRA was asked by government to aug- scenario assuming a continuation of cur- – short-lived low and intermediate level ment this “National Inventory,” with the rent strategy, with respect to spent fuel waste (LILW-SL); radioactivity levels for threefold aim of characterizing extant reprocessing: reprocessing of 79,200 UOX such waste lie as a rule in a range from stocks (state of conditioning, processing fuel assemblies and storage of 5,400 MOX A (next)

MID standard package types Symbols Producers Categories Number Volume (m3) Vitrified waste packages CO — C2 Cogema* HLW 42,470 7,410 Activated metal waste packages B1 EDF ILW-LL 2,560 470 Bituminized sludge packages B2 CEA, Cogema* ILW-LL 105,010 36,060 Cemented technological waste packages B3 CEA, Cogema* ILW-LL 32,940 27,260 Cemented hull and end-cap packages B4 Cogema* ILW-LL 1,520 2,730 Compacted structural and technological waste packages B5 Cogema* ILW-LL 39,900 7,300 Containerized loose structural and technological B6 Cogema* ILW-LL 10,810 4,580 waste packages Total B 192,740 78,400 Total overall 235,210 85,810 * renamed Areva NC in 2006

Table 1. Amounts (number, and volume) of waste packages, as predicted in France for 40 years’ operation of the current fleet of reactors, according to ANDRA’s Dimensioning Inventory Model (MID).

assemblies discharged from the current HLW packages, obtained mainly through materials must thus be precluded, through PWR fleet, when operated over 40 years. vitrification of highly active solutions from the lasting isolation of such waste from the spent fuel reprocessing; environment. This management is guided What forms does it come in? • spent fuel packages: packages consis- by the following principles: to produce as Five types of generic packages (also found ting in nuclear fuel assemblies discharged little waste as practicable; limit its hazar- in MID) may be considered: from reactors; these are not considered to dous character as far as feasible; take into • cementitious waste packages: ILW-LL be waste in France. account the specific characters of each waste packages employing hydraulic-bin- The only long-lived waste packages to be category of waste; and opt for measures der based materials as a conditioning generated in any significant amounts by that will minimize the burden (monitoring, matrix, or as an immobilizing grout, or yet current electricity production (see Box B) maintenance) for future generations. as a container constituent; are vitrified waste packages and standard As for all nuclear activities subject to control • bituminized sludge packages: LLW and compacted waste packages, the other types by the French Nuclear Safety Authority ILW-LL waste packages, in which bitumen of packages having, for the most part, (Autorité de sûreté nucléaire), fundamental is used as confinement matrix for low- and already been produced, and bearing but a safety regulations (RFSs: règles fonda- intermediate-level residues from treat- small part of total radioactivity. mentales de sûreté) have been drawn up ment of a variety of liquid effluents (fuel with respect to radioactive waste mana- processing, research centers, etc.); What is happening to this waste at gement: sorting, volume reduction, pac- • standard compacted waste packages present? What is to be done in the kage confinement potential, manufactu- (CSD-C: colis standard de déchets compac- long term? ring method, radionuclide concentration. tés): ILW-LL packages obtained through The goal of long-term radioactive waste RFS III-2.f, in particular, specifies the condi- compaction conditioning of structural waste management is to protect humankind and tions to be met for the design of, and from fuel assemblies, and technological its environment from the effects of the demonstration of safety for an underground waste from the La Hague workshops; materials comprised in this waste, most repository, and thus provides a basic guide • standard vitrified waste packages importantly from radiological hazards. Any for disposal investigations. Industrial solu- (CSD-V: colis standard de déchets vitrifiés): release or dissemination of radioactive tions (see Industrial solutions for all low- level waste) are currently available for nigh on 85% (by volume) of waste, i.e. VLLW and Short-lived Long-lived LILW-SL waste. A solution for LLW-LL Half-life < 30 years Half-life > 30 years for the main elements waste is the subject of ongoing investigation by ANDRA, at the behest of waste pro- Very-low-level Morvilliers dedicated disposal facility (open since 2003) waste (VLLW) Capacity: 650,000 m3 ducers. ILW-LL and HLW waste, containing radionuclides having very long Low-level waste Dedicated disposal facility under (LLW) investigation for radium-bearing half-lives (in some cases, greater than waste (volume: 100,000 m3) several hundred thousand years) are cur- Aube Center and graphite waste rently held in storage installations coming (open since 1992) (volume: 14,000 m3) Capacity: 1 million m3 under the control of the Nuclear Safety Intermediate-level 3 waste (ILW) MID volume estimate: 78,000 m Authority. What is to become of this waste in the long term, beyond this storage phase, High-level waste MID volume estimate: 7,400 m3 is what the Act of 30 December 1991 (HLW) addresses (see Table 2). Table 2. For all of these waste types, the French Long-term management modes, as currently operated, or planned, in France, by Nuclear Safety Authority is drawing up a radioactive waste category. The orange area highlights those categories targeted by National Radioactive Waste Management investigations covered by the Act of 30 December 1991. Plan, specifying, for each type, a manage- (1) According to the Dimensioning Inventory Model (MID) ment pathway. B Waste from the nuclear power cycle

ost high-level (high-activity) radio- 1 2 Mactive waste (HLW) originates, in H He France, in the irradiation, inside nuclear 3 4 5 6 7 8 9 10 Li Be B C N O F Ne power reactors, of fuel made up from 11 12 13 14 15 16 17 18 enriched uranium oxide (UOX) pellets, or Na Mg Al Si P S Cl Ar also, in part, from mixed uranium and 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 plutonium oxide (MOX). Some 1,200 ton- K Ca Sc Ti V Cr Mn Fe Co Ni Cu Zn Ga Ge As Se Br Kr 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 nes of spent fuel is discharged annually Rb Sr Y Zr Nb Mo Tc Ru Rh Pd Ag Cd In Sn Sb Te I Xe from the fleet of 58 pressurized-water 55 56 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 reactors (PWRs) operated by EDF, sup- Cs Ba Ln Hf Ta W Re Os Ir Pt Au Hg TI Pb Bi Po At Rn 87 88 104 105 106 107 108 109 110 plying over 400 TWh per year, i.e. more Fr Ra An Rf Db Sg Bh Hs Mt Uun than three quarters of French national 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 power consumption. lanthanides La Ce Pr Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb Lu The fuel’s composition alters, during its 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 actinides irradiation inside the reactor. Shortly after Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm Md No Lr discharge, fuel elements contain, on ave- ■ heavy nuclei ■ activation products rage,(1) some 95% residual uranium, 1% ■ fission products ■ fission and activation products plutonium and other transuranic ele- long-lived radionuclides ments – up to 0.1% – and 4% of products yielded by fission. The latter exhibit very Figure 1. significant radioactivity levels – to the The main elements found in spent nuclear fuel. extent this necessitates management safety measures requiring major indus- burnup rate close to 50 GWd/t, result in The plutonium present in spent fuel is trial resources – of some 1017 Bq per tonne bringing down final 235U content to a value yielded by successive neutron capture of initial uranium (tiU) (see Figure 1). quite close to that of natural uranium and decay processes. Part of the Pu is The uranium found in spent fuel exhibits (less than 1%), entailing an energy poten- dissipated through fission: thus about a makeup that is obviously different from tial very close to the latter’s. Indeed, even one third of the energy generated is yiel- that of the initial fuel. The greater the though this uranium remains slightly ded by “in situ recycling” of this element. irradiation, the higher the consumption richer in the fissile isotope than natural These processes further bring about the of fissile nuclei, and consequently the uranium, for which 235U content stands formation of heavy nuclei, involving, whe- greater the extent by which the uranium at 0.7%, the presence should also be ther directly themselves, or through their will have been depleted of the fissile iso- noted, in smaller, though significant, daughter products, long radioactive half- tope 235 (235U). Irradiation conditions amounts, of other isotopes having adverse lives. These are the elements of the acti- usually prevailing in reactors in the French effects in neutronic or radiological terms nide family, this including, essentially, fleet, with an average fuel residence time (232U, 236U), that had not figured in the plutonium (from 238Pu to 242Pu, the odd- inside the reactor of some 4 years, for a initial fuel (see Table 1). numbered isotopes generated in part undergoing fission themselves during (1) These figures should be taken as indicative values. They allow orders of magnitude to be pinpointed for enriched-uranium oxide fuel, taken from the main current French nuclear power irradiation), but equally neptunium (Np), pathway; they do depend, however, on a number of parameters, such as initial fuel composition and americium (Am), and curium (Cm), known irradiation conditions, particularly irradiation time. as minor actinides (MAs), owing to the

UOX 33 GWd/tiU UOX 45 GWd/tiU UOX 60 GWd/tiU MOX 45 GWd/tihm (E 235U: 3.5%) (E 235U: 3.7%) (E 235U: 4.5%) (Ei Pu: 8.65%) element isotope half-life (years) isotope quantity isotope quantity isotope quantity isotope quantity content (g/tiU) content (g/tiU) content (g/tiU) content (g/tihm) (%) (%) (%) (%) 234 246,000 0.02 222 0.02 206 0.02 229 0.02 112 235 7.04·108 1.05 10,300 0.74 6,870 0.62 5,870 0.13 1,070 U 236 2.34·107 0.43 4,224 0.54 4,950 0.66 6,240 0.05 255 238 4.47·109 98.4 941,000 98.7 929,000 98.7 911,000 99.8 886,000 238 87.7 1.8 166 2.9 334 4.5 590 3.9 2,390 239 24,100 58.3 5,680 52.1 5,900 48.9 6,360 37.7 23,100 Pu 240 6,560 22.7 2,214 24,3 2,760 24.5 3,180 32 19,600 241 14.4 12.2 1,187 12.9 1,460 12.6 1,640 14.5 8,920 242 3.75·105 5.0 490 7.8 884 9.5 1,230 11.9 7,300

Table 1. Major actinide inventory for spent UOX and MOX fuel after 3 years’ cooling, for a variety of enrichment and burnup rates. Burnup rate and quantity are expressed per tonne of initial uranium (tiU) for UOX, per tonne of initial heavy metal (tihm) for MOX. B (next)

lesser abundance of these elements, com- UOX 33 GWd/tiU UOX 45 GWd/tiU UOX 60 GWd/tiU MOX 45 GWd/tihm pared with that of U and Pu, the latter being 235 235 235 family (E U: 3.5%) (E U: 3.7%) (E U: 4.5%) (Ei Pu: 8.65%) termed major actinides. quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tihm) Activation processes affecting nuclei of non- radioactive elements mainly involve struc- rare gases (Kr, Xe) 5.6 7.7 10.3 7 tural materials, i.e. the materials of the tubes, grids, plates and end-fittings that alkali metals (Cs, Rb) 345.2 4.5 ensure the mechanical strength of nuclear fuel. These materials lead, in particular, alkaline-earth metals (Sr, Ba) 2.4 3.3 4.5 2.6 to formation of carbon 14 (14C), with a half- Y and life of 5,730 years, in amounts that are lanthanides 10.2 13.8 18.3 12.4 however very low, much less than one gram zirconium 3.6 4.8 6.3 3.3 per tonne of initial uranium (g/tiU) in usual conditions. chalcogens (Se, Te) 0.5 0.7 1 0.8 It is the products yielded by fission of the initial uranium 235, but equally of the Pu molybdenum 3.3 4.5 6 4.1 generated (isotopes 239 and 241), known halogens (I, Br) 0.2 0.3 0.4 0.4 as fission products (FPs), that are the technetium 0.8 1.1 1.4 1.1 essential source of the radioactivity of Ru, Rh, Pd 3.9 5.7 7.7 8.3 spent fuel, shortly after discharge. Over 300 radionuclides – two thirds of which miscellaneous: Ag, Cd, Sn, Sb… 0.1 0.2 0.3 0.6 however will be dissipated through radioactive decay in a few years, after irradia- Table 2. tion – have been identified. These radio- Breakdown by chemical family of fission products in spent UOX and MOX fuel, after 3 years’ cooling, nuclides are distributed over some for a variety of enrichment and burnup rates. 40 elements in the periodic table, from germanium (32Ge) to dysprosium (66Dy), short lifespans, and conversely others For the 500 kg or so of U initially contai- with a presence of tritium from fission, i.e. having radioactive half-lives counted in ned in every fuel element, and after par- from the fission into three fragments (ter- millions of years; and diverse chemical titioning of 475 kg of residual U and about nary fission) of 235U. They are thus cha- properties, as is apparent from the ana- 5 kg Pu, this “ultimate” waste amounts racterized by great diversity: diverse radio- lysis, for the “reference” fuels used in to less than 20 kg of FPs, and less than active properties, involving as they do some PWRs in the French fleet, of the break- 500 grams MAs. This waste management highly radioactive nuclides having very down of FPs generated, by families in the pathway (otherwise know as the closed periodic table (see Table 2). These FPs, cycle), consisting as it does in reproces- along with the actinides generated, are, sing spent fuel now, to partition recove- for the most part, present in the form of rable materials and ultimate waste, dif- oxides included in the initial uranium oxide, fers from strategies whereby spent fuel which remains by far the majority consti- is conserved as-is, whether this be due to tuent. Among some notable exceptions a wait-and-see policy (pending a decision may be noted iodine (I), present in the form on a long-term management mode), or to of cesium iodide, rare gases, such as kryp- a so-called open cycle policy, whereby ton (Kr) and xenon (Xe), or certain noble spent fuel is considered to be waste, and metals, including ruthenium (Ru), rho- designated for conditioning into contai- dium (Rh), and palladium (Pd), which may ners, and disposal as-is. form metallic inclusions within the oxide In the nuclear power cycle, as it is imple- matrix. mented in France, waste is subdivided into Pu is recycled nowadays in the form of two categories, according to its origin. MOX fuel, used in part of the fleet (some Waste directly obtained from spent fuel is 20 reactors currently). Residual U may in further subdivided into minor actinides turn be re-enriched (and recycled as a sub- and fission products, on the one hand, and stitute for mined uranium). Recycling structural waste, comprising hulls (seg- intensity depends on market prices for ments of the cladding tubes that had held natural uranium, the recent upturn in the fuel for PWRs) and end-caps (fittings which should result in raising the current forming the end-pieces of the fuel assem-

Magnum/Harry Gruyaert recycling rate (about one third being recy- blies for these same PWRs), on the other After discharge, spent fuel is stored cled at present). hand. The process used for spent fuel in cooling pools, to allow its radioactivity Such U and Pu recycling is the foundation reprocessing, to extract U and Pu, also to come down significantly. Shown here is a storage pool at Areva’s for the reprocessing strategy currently generates technological waste (operatio- spent fuel reprocessing plant implemented in France, for the major part nal waste, such as spare parts, protec- at La Hague. of spent fuel (some two thirds currently). tion gloves…) and liquid effluents. C What stands between waste and the environment? aw, solid or liquid radioactive waste Rundergoes, after characterization (determination of its chemical and radiological makeup, and of its physical–chemical properties), conditioning, a term covering all the operations consisting in bringing this waste (or spent fuel assemblies) to a form suitable for its transport, storage, and disposal (see Box D). The aim is to put radioactive waste into a solid, physically and chemically stable form, and ensure effective, lasting confinement of the radionuclides it contains. For that purpose, two complementary operations are carried out. As a rule, waste is immobilized by a material – whether by encapsulation or homogeneous incorporation (liquid or powdered A. Gonin/CEA Cross-section of an experimental storage borehole for a spent fuel container (the lower part of the waste, sludges), or encasing (solid waste) assembly may be seen, top right), in the Galatée gallery of CECER (Centre d’expertise sur le – within a matrix, the nature of, and per- conditionnement et l’entreposage des matières radioactives: Radioactive Materials Conditioning formance specification for which depend and Storage Expertise Center), at CEA’s Marcoule Center, showing the nested canisters. on waste type (cement for sludges, evaporation concentrates and incineration ner (cylindrical or rectangular), consis- mary container being the cement or ashes; bitumen for encapsulation of ting in one or more canisters. The whole metal container into which the conditio- sludges or evaporation concentrates – container and content – is termed a ned waste is ultimately placed, to allow from liquid effluent treatment; or a package. Equally, waste may be com- handling. The container may act as initial vitreous matrix, intimately binding the pacted and mechanically immobilized confinement barrier, allotment of func- nuclides to the glass network, for fis- within a canister, the whole forming a tions between matrix and container being sion product or minor actinide solutions). package. determined according to the nature of This matrix contributes to the confine- When in the state they come in as sup- the waste involved. Thus, the whole obtai- ment function. The waste thus conditio- plied by industrial production, they are ned by the grouping together, within one ned is placed in an impervious contai- known as primary packages, the pri- container, of a number of primary C (next) ILW-LL packages may ensure confinement radioactive decay persists inside the pac- Between the containers and the ultimate of the radioactivity of this type of waste. kage (self-irradiationprocess). Emission barrier provided, in a radioactive waste If a long-term storage stage is found to of radiation is concomitant with heat deep disposal facility, by the geological be necessary, beyond the stage of indus- generation. For example, in confinement environment itself, there may further be trial storage on the premises of the pro- glasses holding high-activity (high-level) interposed, apart, possibly, from an over- ducers, primary waste packages must waste, the main sources of irradiation pack, other barriers, so-called engi- be amenable to retrieval, as and when originate in the alpha decay processes neered barriers, for backfill and sealing required: durable primary containers from minor actinides, beta decay from purposes. While these would be point- must then be available, in such condi- fission products, and gamma transitions. less as backfill in clay formations, they tions, for all types of waste. Alpha decay, characterized by produc- would have the capability, in other envi- In such a case, for spent fuel assemblies tion of a recoil nucleus, and emission of ronments (granite), of further retarding which might at some time be earmar- a particle, which, at the end of its path, any flow of radionuclides to the geo- ked for such long-term storage, or even yields a helium atom, causes the major sphere, notwithstanding degradation of for disposal, it is not feasible to demons- part of atom displacements. In particu- the previously mentioned barriers. trate, on a timescale of centuries, the lar, recoil nuclei, shedding considerable integrity of the cladding holding the fuel, energy as they do over a short distance, forming the initial confinement barrier result in atom displacement cascades, during the in-reactor use stage. Securing thus breaking large numbers of chemi- these assemblies in individual, imper- cal bonds. This is thus the main cause vious cartridges is thus being conside- of potential long-term damage. In such red, this stainless-steel cartridge being conditions, matrices must exhibit ther- compatible with the various possible mal stability, and irradiation-damage future management stages: treatment, resistance. return to storage, or disposal. Placing Stored waste packages will also be sub- these cartridges inside impervious jected to the effects of water (leaching). containers ensures a second confine- Container canisters may exhibit a deg- ment barrier, as is the case for high- ree of resistance to corrosion processes level waste packages. (the overpacks contemplated for glas- In storage or disposal conditions, the ses may thus delay by some 4,000 years Technological demonstrators waste packages will be subjected to a the arrival of water), and the confine-

of ILW-LL packages for CEA variety of aggressive agents, both inter- ment matrices must be proven to exhi- bituminized sludges. nal and external. First, radionuclide bit high chemical stability. D From storage to disposal (ILW-LL) in conditions of safety, pen- recycling–transmutation, or simply to he object of nuclear waste storage ding a long-term management mode turn to advantage the natural decay of Tand disposal is to ensure the long- for such waste. Retrieval of stored pac- radioactivity (and hence the falling off term confinement of radioactivity, in kages is anticipated, after a period of of heat release from high-level waste), other words to contain radionuclides limited duration (i.e. after a matter of before putting the waste into geologi- within a definite space, segre- cal disposal. By “long term” is gated from humankind and the meant a timespan of up to 300 environment, as long as requi- years. Long-term storage may red, so that the possible return take place in a surface or sub- to the biosphere of minute surface facility. In the former amounts of radionuclides can case, the site may be protected, have no unacceptable health or for instance, by a reinforced- environmental impact. concrete structure. In the latter According to the Joint Con- case, it will be located at a depth vention on the Safety of Spent of some tens of meters, and pro- Fuel Management and on the tected by a natural environment Safety of Radioactive Waste (for instance, if buried in a hill- Management, signed on 5 Sep- side) and its host rock. tember 1997, “storage” means Whichever management stra- “the holding of spent fuel or of tegy is chosen, it will be impe- radioactive waste in a facility that rative to protect the biosphere provides for its containment, from the residual ultimate waste. with the intention of retrieval.” The nature of the radioelements This is thus, by definition, an the latter contains means a solu- interim stage, amounting to a tion is required that has the abi- delaying, or wait-and-see solu- lity to ensure their confinement tion, even though this may be for over several tens of thousand

a very long time (from a few A. Gonin/CEA years, in the case of long-lived decades to several hundred CEA design study for a common container for waste, or even longer. On such the long-term storage and disposal of long-lived, years), whereas disposal may be intermediate-level waste. timescales, social stability is a final. major uncertainty that has to be Used from the outset of the nuclear years, or tens of years). taken on board. Which is why disposal power age, industrial storage keeps Long-term storage (LTS) may be in deep geological strata (typically, 500 spent fuel awaiting reprocessing, and contemplated, in particular, in the event m down) is seen as a reference solu- conditioned high-level waste (HLW), or of the deferred deployment of a dispo- tion, insofar as it inherently makes for long-lived intermediate-level waste sal facility, or of reactors to carry out deployment of a more passive technical solution, with the ability to stand, with no increased risk, an absence of surveillance, thus mitigating a possible loss of memory on the part of society. excavated diameter: 0.7 m approx. The geological environment of such a disposal facility thus forms a further, length: 40 m approx. essential barrier, which does not exist in the storage case. intercalary A disposal facility may be designed to be reversible over a given period. The lining concept of reversibility means the design disposal packages must guarantee the ability, for a variety of reasons, to access the packages, or even to take them out of the facility, over a certain timespan, or to opt for the final closure of the disposal facility. Such reversibility may be envisaged as a suc- 0.57 m cession of stages, each affording a to 0.64 m decreasing “level of reversibility.” To lid primary package simplify, each stage consists in carrying 1.30m to 1.60 m steel overpack out one further technical operation bringing the facility closer to final closure, ANDRA making retrieval more difficult than at ANDRA design for the disposal of standard vitrified waste packages in horizontal galleries, showing in particular the packages’ various canisters, and some characteristics linked the previous stage, according to well- to potential reversibility of the disposal facility. specified criteria. E What is transmutation?

ransmutation is the transformation of a number of particles, including high- When a neutron collides with a nucleus, Tof one nucleus into another, through energy neutrons. Transmutation by way it may bounce off the nucleus, or pene- a reaction induced by particles with which of direct interaction between protons is trate it. In the latter case, the nucleus, it is bombarded. As applied to the treat- uneconomic, since this would involve, in by absorbing the neutron, gains excess ment of nuclear waste, this consists in order to overcome the Coulomb barrier,(2) energy, which it then releases in various using that type of reaction to transform very-high-energy protons (1–2 GeV), ways: long-lived radioactive isotopes into iso- requiring a generating energy greater • by expelling particles (a neutron, e.g.), topes having a markedly shorter life, or than had been obtained from the process while possibly releasing radiation; even into stable isotopes, in order to that resulted in producing the waste. On • by solely emitting radiation; this is reduce the long-term radiotoxic inven- the other hand, indirect transmutation, known as a capture reaction, since the tory. In theory, the projectiles used may using very-high-energy neutrons (of neutron remains captive inside the be photons, protons, or neutrons. which around 30 may be yielded, depen- nucleus; In the first case, the aim is to obtain, by ding on target nature and incoming proton • by breaking up into two nuclei, of more bremsstrahlung,(1) through bombardment energy), makes it possible to achieve very or less equal size, while releasing concur- of a target by a beam of electrons, pro- significantly improved performance. This rently two or three neutrons; this is known vided by an accelerator, photons able to is the path forming the basis for the as a fission reaction, in which considera- bring about reactions of the (γ, xn) type. design of so-called hybrid reactors, cou- ble amounts of energy are released. Under the effects of the incoming gamma pling a subcritical core and a high-inten- Transmutation of a radionuclide may be radiation, x neutrons are expelled from sity proton accelerator (see Box F, What achieved either through neutron capture the nucleus. When applied to substan- is an ADS?). or by fission. Minor actinides, as elements ces that are too rich in neutrons, and The third particle that may be used is thus having large nuclei (heavy nuclei), may hence unstable, such as certain fission the neutron. Owing to its lack of electric undergo both fission and capture reac- products (strontium 90, cesium 137…), charge, this is by far the particle best sui- tions. By fission, they transform into such reactions yield, as a rule, stable sub- ted to meet the desired criteria. It is “natu- radionuclides that, in a majority of cases, stances. However, owing to the very low rally” available in large quantities inside are short-lived, or even into stable nuclei. efficiency achieved, and the very high nuclear reactors, where it is used to trig- The nuclei yielded by fission (known as electron current intensity required, this ger fission reactions, thus yielding energy, fission products), being smaller, are only path is not deemed to be viable. while constantly inducing, concurrently, the seat of capture reactions, undergoing, In the second case, the proton–nucleus transmutations, most of them unsought. on average, 4 radioactive decays, with a interaction induces a complex reaction, The best recycling path for waste would half-life not longer than a few years, as known as spallation, resulting in frag- thus be to reinject it in the very installa- a rule, before they reach a stable form. mentation of the nucleus, and the release tion, more or less, that had produced it… Through capture, the same heavy nuclei (1) From the German for “braking radiation.”High-energy photon radiation, yielded by accelerated transform into other radionuclides, often (or decelerated) particles (electrons) following a circular path, at the same time emitting braking long-lived, which transform in turn photons tangentially, those with the highest energies being emitted preferentially along the electron through natural decay, but equally beam axis. through capture and fission. (2) A force of repulsion, which resists the drawing together of same-sign electric charges. E (next)

The probability, for a neutron, of causing tion prevails; capture is about 100 times mal neutron pathway is the technology a capture or a fission reaction is evalua- more probable than fission. This remains used by France for its power generation ted on the basis, respectively, of its cap- the case for energies in the 1 eV–1 MeV equipment, with close to 60 pressurized- ture cross-section and fission cross-sec- range (i.e., that of epithermal neutrons, water reactors. In a thermal-neutron tion. Such cross-sections depend on the where captures or fissions occur at defi- reactor, neutrons yielded by fission are nature of the nucleus (they vary consi- nite energy levels). Beyond 1 MeV (fast slowed down (moderated) through colli- derably from one nucleus to the next, and, neutron range), fissions become more sions against light nuclei, making up even more markedly, from one isotope probable than captures. materials known as moderators. Due to to the next for the same nucleus) and Two reactor pathways may be conside- the moderator (common water, in the neutron energy. red, according to the neutron energy case of pressurized-water reactors), neu- For a neutron having an energy lower range for which the majority of fission tron velocity falls off, down to a few kilo- than 1 eV (in the range of slow, or ther- reactions occur: thermal-neutron reac- meters per second, a value at which neu- mal, neutrons), the capture cross-sec- tors, and fast-neutron reactors. The ther- trons find themselves in thermal equilibrium with the ambient environ- capture ment. Since fission cross-sections for 242 243Cm 235U and 239Pu, for fission induced by fission Cm 4% 11% 163 32 thermal neutrons, are very large, a days years beta-minus 95% concentration of a few per cent of these decay 3% 1% 86% fissile nuclei is sufficient to sustain the alpha cascade of fissions. The flux, in a ther- 86% decay mal-neutron reactor, is of the order of 241Am 86% 242Am 242mAm 15% 1018 neutrons per square meter, per electron 432 16 141 capture ans 12% h years second. 2% 14% 85% In a fast-neutron reactor, such as Phénix, neutrons yielded by fission immediately 7% induce, without first being slowed down, 238Pu 239Pu 240Pu 241Pu 87.8 82% 2.4.104 36% 6.5.103 98% 14.4 24% further fissions. There is no moderator in 5% years years years years this case. Since, for this energy range, 13% 64% 2% 69% cross-sections are small, a fuel rich in fissile radionuclides must be used (up to 20% uranium 235 or plutonium 239), if the Figure. Simplified representation of the evolution chain of americium 241 in a thermal-neutron reactor neutron multiplication factor is to be equal (shown in blue: radionuclides disappearing through fission). Through capture, 241Am transforms to 1. The flux in a fast-neutron reactor is into 242mAm, this disappearing predominantly through fission, and into 242Am, which mainly decays ten times larger (of the order of 1019 neu- (with a half-life of 16 hours) through beta decay into 242Cm. 242Cm transforms through alpha decay into 238Pu, and through capture into 243Cm, which itself disappears predominantly through fission. trons per square meter, per second) than 238Pu transforms through capture into 239Pu, which disappears predominantly through fission. for a thermal-neutron reactor. F What is an ADS?

n ADS (accelerator-driven system) is for a reactor in normal operating condi- ter then go on to interact with the fuel of Aa hybrid system, comprising a tions; and, if keff is lower than 1, the neu- the subcritical neutron multiplier nuclear reactor operating in subcritical tron population dwindles, and becomes medium, yielding further neutrons (fis- mode, i.e. a reactor unable by itself to sus- extinct, unless – as is the case for a hybrid sion neutrons) (see Figure). tain a fission chain reaction, “driven” by system – an external source provides a Most hybrid system projects use as a core an external source, having the ability to neutron supply. (of annular configuration, as a rule) fast- supply it with the required comple- neutron environments, since these ment of neutrons.(1) make it possible to achieve neu- Inside the core of a nuclear reactor, window tron balances most favorable to indeed, it is the fission energy from transmutation, an operation that accelerator spallation heavy nuclei, such as uranium 235 target allows waste to be “burned,” but or plutonium 239, that is released. providing which may equally be used to yield Uranium 235 yields, when under- 100 keV 1 GeV external further fissile nuclei. Such a sys- neutrons going fission, on average 2.5 neu- proton tem may also be used for energy source trons, which can in turn induce a subcritical reactor generation, even though part of further fission, if they collide with a this energy must be set aside to uranium 235 nucleus. It may thus power the proton accelerator, a be seen that, once the initial fission Principle schematic of an ADS. part that is all the higher, the more is initiated, a chain reaction may develop, From the effective multiplication factor, subcritical the system is. Such a system resulting, through a succession of fis- a reactor’s reactivity is defined by the ratio is safe in principle from most reactivity

sions, in a rise in the neutron population. (keff –1)/keff. The condition for stability is accidents, its multiplication factor being However, of the 2.5 neutrons yielded by then expressed by zero reactivity. To sta- lower than 1, contrary to that of a reac- the initial fission, some are captured, thus bilize a neutron population, it is sufficient tor operated in critical mode: the chain not giving rise to further fissions. The to act on the proportion of materials exhi- reaction would come to a halt, if it was number of fissions generated from one biting a large neutron capture cross-sec- not sustained by this supply of external initial fission is characterized by the effec- tion (neutron absorber materials) inside neutrons.

tive multiplication factor keff, equal to the the reactor. A major component in a hybrid reactor, ratio of the number of fission neutrons In an ADS, the source of extra neutrons the window, positioned at the end of the generated, over the number of neutrons is fed with protons, generated with an beam line, isolates the accelerator from disappearing. It is on the value of this coef- energy of about 100 keV, then injected the target, and makes it possible to keep ficient that the evolution of the neutron into an accelerator (linear accelerator or the accelerator in a vacuum. Traversed

population depends: if keff is markedly cyclotron), which brings them to an energy as it is by the proton beam, it is a sensi- higher than 1, the population increases of around 1 GeV, and directs them to a tive part of the system: its lifespan rapidly; if it is slightly higher than 1, neu- heavy-metal target (lead, lead–bismuth, depends on thermal and mechanical tron multiplication sets in, but remains tungsten or tantalum). When irradiated stresses, and corrosion. Projects are moo- under control; this is the state desired at by the proton beam, this target yields, ted, however, of windowless ADSs. In the

reactor startup; if keff is equal to 1, the through spallation reactions, an intense, latter case, it is the confinement cons- population remains stable; this is the state high-energy (1–20 MeV) neutron flux, one traints, and those of radioactive spalla- single incoming neutron having the abi- tion product extraction, that must be taken (1) On this topic, see Clefs CEA,No.37, p. 14 lity to generate up to 30 neutrons. The lat- on board. The industrial context 1

The characteristics of the major part of the radioactive waste generated in France are determined by those of the French nuclear power generation fleet, and of the spent fuel reprocessing plants, built in compliance with the principle of reprocessing such fuel, to partition such materials as remain recoverable for energy purposes (uraniumand plutonium), and waste (fission productsand minor actinides), not amenable to recycling in the current state of the art. 58 enriched-uranium pressurized-water reactors (PWRs) have been put on stream by French national utility EDF, from 1977 (Fessenheim) to 1999 (Civaux), forming a second generation of reactors, following the first generation, which mainly comprised 8 UNGG (natural uranium, graphite, gas) reactors, now all closed down, and, in the case of the older reactors, in the course of decommis- sioning. Some 20 of these PWRs carry out the industrial recycling of plutonium, included in MOX fuel, supplied since 1995 by the Melox plant, at Marcoule (Gard département, Southern France). EDF is contemplating the gradual replacement of the current PWRs by third-generation reactors, belonging to the selfsame pressurized-water reactor pathway, of the EPR (European Pressurized-Water Reactor) type, designed by Areva NP (formerly Framatome–ANP), a division of the Areva Group. The very first EPR is being built in Finland, the first to be built in France being sited at Flamanville (Manche département, Western France). The major part of spent fuel from the French fleet currently undergoes reprocessing at the UP2-800(1) plant, which has been operated at La Hague (Manche département), since 1994, by Areva NC (formerly Cogema,) another member of the Areva Group (the UP3 plant, put on stream in 1990–92, for its part, carries out reprocessing of fuel from other countries). The waste vitrification workshops at these plants, the outcome of development work initiated at Marcoule, give their name (R7T7) to the “nuclear” glass used for the confinement of long-lived, high-level waste. A fourth generation of reactors could emerge from 2040 (along with new reprocessing plants), a prototype being built by 2020. These could be fast-neutron reactors (i.e. fast reactors [FRs]), either sodium-cooled (SFRs) or gas-cooled (GFRs). Following the closing down of the Superphénix reactor, in 1998, only one FR is operated in France, the Phénix reactor, due to be closed down in 2009.

(1) A reengineering of the UP2-400 plant, which, after the UP1 plant, at Marcoule, had been intended to reprocess spent fuel from the UNGG pathway.