spent fuel reprocessing: a fully mastered pathway · (1) these figures should be taken as...
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CLEFS CEA - No. 53 - WINTER 2005-2006 19
Spent fuel reprocessing:a fully mastered pathway
Spent fuel reprocessing, as carried out in France, seeks to optimize both the partitioning of recoverable materials, and management of ultimate waste. The current process used for spent fuel reprocessing, drawing on over 50 years’ research and industrial development in a number of countries, allows extraction, for recycling purpose, of uranium and plutonium,with very high recovery and purification rates.
Irradiation, inside a reactor, of a fuel assembly generates, through nuclear reactions, a widevariety of radionuclides, within the fuel and in the metal components making up theassembly’s structure.
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So-called “spent”nuclear fuel,discharged on a regu-lar basis from nuclear power reactors, forms the
overwhelming part in the “source term”for long-lived,high-activity nuclear waste (HLW-LL). The originalfuel – uranium (U),most commonly – has undergonefar-reaching transformations, due to the nuclear fis-sion and capture reactions taking place in it, in thecourse of its few years’residence inside the reactor,andit is a highly “variegated” assembly that is thus takenout: residual uranium indeed as majority constituent,along however with plutonium (Pu) and other trans-uranic elements, fission products (FPs) and activa-tion products (APs)… (see Box B, Waste from thenuclear power cycle).
What strategy to adopt regarding spent fuel?
The various constituents of spent nuclear fuel exhibithighly diverse physical or chemical properties.However,two main characteristics govern the various optionsfor spent fuel management. On the one hand, over-whelmingly, the nuclei involved are radioactive nuclei,with a very broad spread of radioactive half-lives (ran-ging from very short to very long). They thus repre-sent a potential hazard. And, on the other hand, someof these nuclei are fissile, as are the major part of Pu
Some 1,200 tonnes of spent fuel is discharged annually fromthe 58 pressurized-water reactors (PWRs) in the Frenchnuclear power plant fleet, this generating over 400 TWh/year.Shown here, the Civaux (Vienne département, WesternFrance) power plant, comprising two 1,450-MWe units.
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The groundwork of current solutions
B
Most high-level (high-activity) radio-active waste (HLW) originates, in
France, in the irradiation, inside nuclearpower reactors, of fuel made up fromenriched uranium oxide (UOX) pellets, oralso, in part, from mixed uranium andplutonium oxide (MOX). Some 1,200 ton-nes of spent fuel is discharged annuallyfrom the fleet of 58 pressurized-waterreactors (PWRs) operated by EDF, sup-plying over 400 TWh per year, i.e. morethan three quarters of French nationalpower consumption.The fuel’s composition alters, during itsirradiation inside the reactor. Shortly afterdischarge, fuel elements contain, on ave-rage,(1) some 95% residual uranium, 1%plutonium and other transuranic ele-ments – up to 0.1% – and 4% of productsyielded by fission. The latter exhibit verysignificant radioactivity levels – to theextent this necessitates managementsafety measures requiring major indus-trial resources – of some 1017 Bq per tonneof initial uranium (tiU) (see Figure 1).The uranium found in spent fuel exhibitsa makeup that is obviously different fromthat of the initial fuel. The greater theirradiation, the higher the consumptionof fissile nuclei, and consequently thegreater the extent by which the uraniumwill have been depleted of the fissile iso-tope 235 (235U). Irradiation conditionsusually prevailing in reactors in the Frenchfleet, with an average fuel residence timeinside the reactor of some 4 years, for a
burnup rate close to 50 GWd/t, result inbringing down final 235U content to a valuequite close to that of natural uranium(less than 1%), entailing an energy poten-tial very close to the latter’s. Indeed, eventhough this uranium remains slightlyricher in the fissile isotope than naturaluranium, for which 235U content standsat 0.7%, the presence should also benoted, in smaller, though significant,amounts, of other isotopes having adverseeffects in neutronic or radiological terms(232U, 236U), that had not figured in theinitial fuel (see Table 1).
The plutonium present in spent fuel isyielded by successive neutron captureand decay processes. Part of the Pu isdissipated through fission: thus aboutone third of the energy generated is yiel-ded by “in situ recycling” of this element.These processes further bring about theformation of heavy nuclei, involving, whe-ther directly themselves, or through theirdaughter products, long radioactive half-lives. These are the elements of the acti-nide family, this including, essentially,plutonium (from 238Pu to 242Pu, the odd-numbered isotopes generated in partundergoing fission themselves duringirradiation), but equally neptunium (Np),americium (Am), and curium (Cm), knownas minor actinides (MAs), owing to the
Waste from the nuclear power cycle
(1) These figures should be taken as indicative values. They allow orders of magnitude to bepinpointed for enriched-uranium oxide fuel, taken from the main current French nuclear powerpathway; they do depend, however, on a number of parameters, such as initial fuel composition andirradiation conditions, particularly irradiation time.
Figure 1.The main elements found in spent nuclear fuel.
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 1,070U
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
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%)
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 areexpressed per tonne of initial uranium (tiU) for UOX, per tonne of initial heavy metal (tihm) for MOX.
H1
Li3
Na11
K19
Rb37
Cs55
Fr87
Be4
Mg12
Ca20
Sr38
Ba56
Ra88
Sc21
Y39
Ln
An
Ti22
Zr40
Hf72
Rf104
V23
Nb41
Ta73
Db105
Cr24
Mo42
W74
Sg106
Mn25
Tc43
Re75
Bh107
Fe26
Ru44
Os76
Hs108
Co27
Rh45
Ir77
Mt109
Ni28
Pd46
Pt78
Uun110
Ac89
Th90
Pa91
U92
Np93
Pu94
Am95
Cm96
Bk97
Cf98
Es99
Fm100
Md101
No102
Lr103
Cu29
Ag47
Au79
Zn30
Cd48
Hg80
Ga31
In49
TI81
Ge32
Sn50
Pb82
As33
Sb51
Bi83
Se34
Te52
Po84
Br35
I53
At85
Kr36
Al13
Si14
P15
S16
Cl17
Ar18
B5
C6
N7
O8
F9
Ne10
He2
Xe54
Rn86
La57
Ce58
Pr59
Nd60
Pm61
Sm62
Eu63
Gd64
Tb65
Dy66
Ho67
Er68
Tm69
Yb70
Lu71
■ heavy nuclei■ fission products
long-lived radionuclides
■ activation products■ fission and activation products
actinides
lanthanides
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lesser abundance of these elements, com-pared with that of U and Pu, the latter beingtermed major actinides.Activation processes affecting nuclei of non-radioactive elements mainly involve struc-tural materials, i.e. the materials of thetubes, grids, plates and end-fittings thatensure the mechanical strength of nuclearfuel. These materials lead, in particular,to formation of carbon 14 (14C), with a half-life of 5,730 years, in amounts that arehowever very low, much less than one gramper tonne of initial uranium (g/tiU) in usualconditions.It is the products yielded by fission of theinitial uranium 235, but equally of the Pugenerated (isotopes 239 and 241), knownas fission products (FPs), that are theessential source of the radioactivity ofspent fuel, shortly after discharge. Over300 radionuclides – two thirds of whichhowever will be dissipated through radio-active decay in a few years, after irradia-tion – have been identified. These radio-nuclides are distributed over some40 elements in the periodic table, fromgermanium (32Ge) to dysprosium (66Dy),with a presence of tritium from fission, i.e.from the fission into three fragments (ter-nary fission) of 235U. They are thus cha-racterized by great diversity: diverse radio-active properties, involving as they do somehighly radioactive nuclides having very
short lifespans, and conversely othershaving radioactive half-lives counted inmillions of years; and diverse chemicalproperties, as is apparent from the ana-lysis, for the “reference” fuels used inPWRs in the French fleet, of the break-down of FPs generated, by families in theperiodic table (see Table 2). These FPs,along with the actinides generated, are,for the most part, present in the form ofoxides included in the initial uranium oxide,which remains by far the majority consti-tuent. Among some notable exceptionsmay be noted iodine (I), present in the formof cesium iodide, rare gases, such as kryp-ton (Kr) and xenon (Xe), or certain noblemetals, including ruthenium (Ru), rho-dium (Rh), and palladium (Pd), which mayform metallic inclusions within the oxidematrix.Pu is recycled nowadays in the form ofMOX fuel, used in part of the fleet (some20 reactors currently). Residual U may inturn be re-enriched (and recycled as a sub-stitute for mined uranium). Recyclingintensity depends on market prices fornatural uranium, the recent upturn inwhich should result in raising the currentrecycling rate (about one third being recy-cled at present).Such U and Pu recycling is the foundationfor the reprocessing strategy currentlyimplemented in France, for the major partof spent fuel (some two thirds currently).
For the 500 kg or so of U initially contai-ned in every fuel element, and after par-titioning of 475 kg of residual U and about5 kg Pu, this “ultimate” waste amountsto less than 20 kg of FPs, and less than500 grams MAs. This waste managementpathway (otherwise know as the closedcycle), consisting as it does in reproces-sing spent fuel now, to partition recove-rable materials and ultimate waste, dif-fers from strategies whereby spent fuelis conserved as-is, whether this be due toa wait-and-see policy (pending a decisionon a long-term management mode), or toa so-called open cycle policy, wherebyspent fuel is considered to be waste, anddesignated for conditioning into contai-ners, and disposal as-is.In the nuclear power cycle, as it is imple-mented in France, waste is subdivided intotwo categories, according to its origin.Waste directly obtained from spent fuel isfurther subdivided into minor actinidesand fission products, on the one hand, andstructural waste, comprising hulls (seg-ments of the cladding tubes that had heldthe fuel for PWRs) and end-caps (fittingsforming the end-pieces of the fuel assem-blies for these same PWRs), on the otherhand. The process used for spent fuelreprocessing, to extract U and Pu, alsogenerates technological waste (operatio-nal waste, such as spare parts, protec-tion gloves…) and liquid effluents.
UOX 33 GWd/tiU UOX 45 GWd/tiU UOX 60 GWd/tiU MOX 45 GWd/tihmfamily (E 235U: 3.5%) (E 235U: 3.7%) (E 235U: 4.5%) (Ei Pu: 8.65%)
quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tihm)
rare gases (Kr, Xe) 5.6 7.7 10.3 7
alkali metals(Cs, Rb) 3 4 5.2 4.5
alkaline-earthmetals (Sr, Ba) 2.4 3.3 4.5 2.6
Y and lanthanides 10.2 13.8 18.3 12.4
zirconium 3.6 4.8 6.3 3.3
chalcogens(Se, Te) 0.5 0.7 1 0.8
molybdenum 3.3 4.5 6 4.1
halogens (I, Br) 0.2 0.3 0.4 0.4
technetium 0.8 1.1 1.4 1.1
Ru, Rh, Pd 3.9 5.7 7.7 8.3
miscellaneous:Ag, Cd, Sn, Sb… 0.1 0.2 0.3 0.6
Table 2.Breakdown by chemical family of fission products in spent UOX and MOX fuel, after 3 years’ cooling,for a variety of enrichment and burnup rates.
After discharge, spent fuel is stored in cooling pools, to allow its radioactivity to come down significantly. Shown here is a storage pool at Areva’s spent fuel reprocessing plant at La Hague.
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The groundwork of current solutions
isotopes, or fertile, as is the case for most isotopes ofthe residual U. They thus present an energy potentialof interest,which may yet be put to use in nuclear reac-tors.This dual character – potential resource, and noxiousagent – of spent fuel means a conditioning and recy-cling strategy makes sense, aiming for the discrimi-nating management of waste, conditioning it intodurable forms, and of the (majority) recoverable ele-ments, through partitioning and recycling. Spent
nuclear fuel recycling, as currently carried out inFrance, is part and parcel of just such a strategy (seeFigure 1). Pu extraction and recycling, in the formof MOX fuel, and that of so-called “reprocessed” Uare nowadays an industrial reality. FPs and othertransuranic elements are, for the most part, concen-trated and conditioned in the form of glass blocks,which have been shown to have the ability to pro-vide effective confinement over very long timespans.Such a scheme may be seen as the implementation,for the nuclear sector, of the more general philoso-phy of conserving natural resources, and limitingenvironmental impact, through systematic recycling.This, however, is not taken up by all countries ope-rating nuclear power plant fleets. Some of these, suchas the United States or Sweden, have rather opted fordisposal of spent fuel as-is, by deference to a num-ber of considerations, be they economic, technical,political… Japan, as France, has resolutely gone forthe recycling path. Other countries, such as China,are seeking to follow suit. Finally, many countries,e.g. Spain or Taiwan, are presently keeping their spentfuel in storage, pending a decision on long-term stra-tegy.Be that as it may, analysis of the challenges for thelonger term clearly shows that sustainable nuclearpower options do involve recycling, as evidenced bythe work of the Generation IV International Forum,which is looking into future nuclear power systemscoming into consideration at the turn of the cen-tury.
What is the process currently used for spent fuel reprocessing?
The process used for spent fuel reprocessing is cur-rently,virtually to the exclusion of all others, the Purexprocess. Based on selective U and Pu extraction, bymeans of an organic compound, tributyl phosphate(TBP), it can now avail itself of half a century ofresearch and development, since the concept was wor-ked out in the 1940s in the United States. By the fol-lowing decade, it had supplanted the various otherpathways that had been investigated for Pu recoveryfrom irradiated materials, leading to industrial appli-
Figure 1.Principle schematic of
reprocessing–recycling.Reprocessing of spent
UOX fuel allows uraniumand plutonium
to be recovered for recycling purposes,the former in the form
of UOX fuel after re-enrichment,
the latter in the form of MOX fuel.
CEA
A nuclear fuel assembly.About 4 m long,
with a 9.5-mm-diametersection, each
of the 264 fuel rods, held together by grids
to form an assembly,comprises a cladding
tube, 580 µm thick, in Zircaloy®
(a zirconium-based metal alloy), into which
265 fuel pellets arestacked.
spiderassembly
control rod
hold-downspring
upper end-fitting
upper grid
thimble tube
fuel rod
mixing grid
lower grid
lower end-fitting
waste
reactor
reprocessing
uranium enrichment
mining and refining activities
fuel fabrication
uranium
plutonium
(UOX)
(MOX)spent fuel
CLEFS CEA - No. 53 - WINTER 2005-2006 23
cations in the United States (Savannah River, 1954),the United Kingdom (Windscale, 1964), and France(Marcoule, UP1 Plant, 1958). It is used by Areva atLa Hague (in a plant with the design capacity to servereactors generating over 400 TWh/year), by BNFL inthe United Kingdom (THORP [Thermal OxideReprocessing Plant] facility, on the Sellafield [for-
merly Windscale] site), and at a similar unit startingup at Rokkasho-Mura (Japan).The process consists, initially, in dissolving all of theconstituent elements of the spent fuel,going on to effectselective U and Pu extraction from this solution, takingadvantage of the selective affinity exhibited by TBP forthese elements (see Figure 2). These operations, com-plicated as they are owing to the variety of species pre-sent, but equally due to the specific conditions requi-red to set up the operations, and the strict constraintsthat must be taken on board, allow very high partitio-ning rates to be achieved. U and Pu are recovered tobetter than 99.8%,ultimately accounting for only verylow residual contamination (purification factors of theorder of 107), for recycling purposes. The activationproducts (APs), minor actinides (MAs) and fission
products (FPs) making up the waste may then be condi-tioned as-is, using appropriate processes.
What are the main elementary operations?Reprocessing operations are carried out after a periodof storage, currently of a few years or so, on average.This stage allows significant decay of the spent fuel’sradioactivity,and thus equally a downgrading of ther-mal power release. Thus, three years after discharge,residual power and activity levels, for one tonne ofinitial uranium, stand at a few kilowatts (kW), andsome 1013 Bq, respectively.“Head-end” operations consist, subsequent to dis-mantling of the fuel assemblies, in the successiveshearing, then dissolution of the spent fuel, going onto the processing of gases yielded by these steps. Thefuel rods, holding enriched uranium pellets, arechopped into segments a few centimeters long bymeans of shears, to “lay bare” the fuel. The latter isthen dissolved in an aqueous solution of concentra-ted nitric acid (HNO3). U and Pu are thus dissolvedin the form of nitrates, in accordance with the mainreactions, as follows:
UO2 + 3 HNO3 p UO2+2 + 2 NO-
3 + 1/2 NO + 1/2 NO2
+ 3/2 H2O
PuO2 + 4 HNO3p Pu4+ + 4 NO-3 + 2 H2O.
Battery of mixer-settler equipment, comprising 12 units set up in series, used from 1982 to 1994 to partition uranium and plutonium streams during spent fuelreprocessing.
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Figure 2.Principle schematic of the Purex process. After dissolution in nitricacid (HNO3) of the irradiatedfuel, uranium and plutoniumare selectively extractedfrom the solution throughuse of TBP.
Pulsed column used in the Purex process, at Areva’sUP3 plant at La Hague, for uranium and plutoniumextraction.
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dissolution
irradiated fuel HNO3
TBP
extractionsolution
U, Pu, FPs, MAs
hulls FPs + MAs
U
Pu
This dissolution does not affect the fuel cladding ele-ments and other structural items (hulls and end-caps), which, owing to their composition (stainlesssteel, Zircaloy®…) remain unaltered during this ope-ration. A fraction of certain elements (platinoids,technetium) is resistant to this dissolution process,leading to the presence in the resulting solution ofsolid compounds (sub-micron particulates, for themost part). These are usually separated out throughconventional settling (gravity or centrifuge) or fil-tration techniques, and the residue thus obtainedamounts, in mass terms, to a few parts per thousandor so. Shearing, then acid attack of the fuel releasegaseous FPs: krypton (Kr), xenon (Xe), but equallyiodine(1) (I) – this being formed and desorbed as aresult of the dissolution of cesium iodide – or CO2,owing to the presence of carbon 14. These opera-tions further yield nitrogen oxides (NO and NO2),
radioactive aerosols, and tritiated water vapor. Thisgas stream is subjected to scrubbing, to retain entrai-ned particulates, allow recombination of nitrogenoxides into nitric acid, and induce a backflow of tri-tium (3H). Iodine is trapped in the form of iodidesin an alkaline solution, or on solid sorbents.Regarding rare gases, only 85Kr is radioactive. As for133Xe, another highly abundant nuclide, this is a sta-ble isotope.The Purex process allows, through the selective affi-nity of the TBP molecule, extraction of U and Pu,which are hexavalent and tetravalent, respectively,without however extracting, as a whole, MAs andFPs, these being trivalent. Extraction of the formerelements is reversible, inasmuch as it is sufficient toalter certain chemical conditions to achieve eitherextraction or back-extraction (the reverse process)of U or Pu. For example, lower medium nitratecontent, or, for Pu, reduction of the metal to oxi-dation state + III, exhibiting very poor affinity forTBP, allows the process to be reversed, thus retur-ning to aqueous solution the species previously extrac-ted. The Purex process thus makes use of successiveU and Pu extraction and back-extraction cycles (seeFigure 3), to achieve quantity recovery, and purifi-cation of these elements.Aside from the diversity of the species involved, afurther complicating factor is the prevalence, through-out the operations, of a significant radioactivity level,indeed a high level for some operations. The radio-sensitivity of certain species, particularly that of theextractant, even if it be low, must be taken on board.Complementary separation operations are carriedout, in particular, to segregate the byproducts gene-rated, and enable recycling of the reactants involved.Such recycling, given the strict requirement to curbsecondary waste, is one of the requisites, for the pro-cess’s industrial viability.
What happens to the recycled materials and waste?The Purex process thus makes it possible, through asuccession of mechanical, then chemical operations,
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The groundwork of current solutions
Figure 3.Principle schematic of an extraction cycle in the Purex process.Successive extractionand back-extractioncycles enable, throughuse of TBP, uraniumand plutonium recoveryand purification.
After extraction of the uranium and plutonium,
waste comprising all minor actinides
and nearly all fissionproducts is conditioned
by integrating it into a glass matrix.
Seen here is the storagehall of vitrification
workshop R7 at Areva’sUP2-800 spent fuelreprocessing plant,
at La Hague. Are
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(1) It should be noted that the only isotopes of iodine that need be considered, in reprocessing operations, are isotopes 127 (stable)and 129 (which has a very long half-life). Isotope 131,which is highly radioactive, generated as it is during irradiation,and present in significant quantities in in-reactor fuel,has promptly vanished during the storage stage, owing to its shorthalf-life (8 days).
U, Pu U
UPu
dilute TBPaqueous solution
FPs, MAs
extraction Puback-extraction
Uback-extraction
U, Pu, FP, MAfeed
extractantregeneration
CLEFS CEA - No. 53 - WINTER 2005-2006 25
to achieve the intended separations. At the end of theprocess, two types of products are obtained (seeFigure 4).Recyclable materials, U and Pu, for which purity levelsare very high, owing to the very high selectivity ofTBP for the elements to be partitioned, go on to beprocessed into a form meeting the specifications ofrecycling (e.g.plutonium dioxide PuO2, for the manu-facture of MOX fuel).Waste, as obtained from spent fuel, comes in twoforms.All of the MAs and nearly all of the FPs initiallypresent in the spent fuel are brought together intoone single solution, after merging of the variousstreams issuing from the successive extraction cycles.Currently, the technique used for this solution consists,as a rule, in carrying out a concentration operation,and then seek to produce a particularly stable condi-tioning. In the three countries (France, the UnitedKingdom and Japan) operating the Purex process onan industrial scale, the vitrification process is used.Structural waste contains the major part of APs gene-rated during irradiation, accounting for less than 1%of radioactivity in irradiated fuel, a few years after ithas been discharged. Aside from these three mainproducts, mention should be made of technologi-cal waste, not representing any large amounts, asso-ciated to operation of the process. This waste is condi-tioned in suitable, durable form by the La Haguereprocessing plants (see Industrial solutions for long-lived, high- and intermediate-level waste).The technique of spent fuel reprocessing is well inhand nowadays. The Purex process stands as theworld reference, in industrial terms, and yields remar-kable performance, most conspicuously in terms ofrecovery rates for recyclable materials, and of volume,quality of confinement and long-term radiotoxicityas regards ultimate waste. Recycling U and Pu mayenable savings of up to 30% of the initial uranium,with current reactor technology.One of the goals set for the research effort carriedout under the aegis of the French Act of 30 December1991 (see Box 2 in Radioactive waste management
Figure 4.Streamsgenerated duringoperation of thePurex process(operationalfindings at theAreva plants at La Hague).
MOX fuel rods. Plutonium yielded by reprocessing spent UOX fuel from the EDF reactor fleet is recycled in the form of MOX fuel on an industrial scale, used in 20 reactors in the fleet.
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research: an ongoing process of advances) is, after stri-ving – successfully – to achieve lower volumes ofwaste from these operations, that of investigatingpaths for a reduction of its long-term toxicity.
> Bernard BoullisNuclear Energy Division
CEA Valrhô-Marcoule Center> Jean-Guy Devezeaux de Lavergne
Treatment Business UnitAreva
hulls and end-capsglasses glasses
liquid effluents
3H: ~ 10 TBq
uranium
950 kg
� emitters : < 0.1 GBq� emitters : < 0.1 GBq
gaseous effluents
Kr: ~ 300 TBq
plutonium
~ 9.56 kg
� emitters : < 0.1 GBq
Purex
1 t irradiated uranium
U : 955 kgPu : 9.7 kgFPs : 34 kgMAs : 750 g
� emitters : 100 TBq� emitters : 26,000 TBq
(APs) : 200 TBq waste
A
According to the International AtomicEnergy Agency (IAEA), radioactive
waste may be defined as “any material forwhich no use is foreseen and that containsradionuclides at concentrations greaterthan the values deemed admissible by thecompetent authority in materials suitablefor use not subject to control.” French lawin turn introduces a further distinction,valid for nuclear waste as for any otherwaste, between waste and final, or “ulti-mate,” waste (déchet ultime). Article L.541-1 of the French Environmental Codethus specifies that “may be deemed aswaste any residue from a process of pro-duction, transformation or use, any sub-stance, material, product, or, more gene-rally, any movable property left derelict orthat its owner intends to leave derelict,”further defining as ultimate “waste, be itthe outcome of waste treatment or not,that is not amenable to further treatmentunder prevailing technological and eco-nomic conditions, in particular by extrac-tion of the recoverable, usable part, ormitigation of its polluting or hazardouscharacter.”Internationally, experts from IAEA and theNuclear Energy Agency (NEA) – an OECDorganization – as those in the EuropeanCommission find that long-lived waste pro-duced in countries operating a nuclearpower program is stored securely nowa-days, whilst acknowledging a final solu-tion is required, for the long-term mana-gement of such waste. They consider burialin deep geological structures appears, pre-sently, to be the safest way to achieve finaldisposal of this type of waste.
What constitutes radioactivewaste? What are the volumescurrently involved?Radioactive waste is classified into a num-ber of categories, according to its level ofradioactivity, and the radioactive period,or half-life, of the radionuclides itcontains. It is termed long-lived wastewhen that period is greater than 30 years,short-lived waste otherwise. The Frenchclassification system involves the follo-wing categories:– very-low-level waste (VLLW); thiscontains very small amounts of radionu-clides, of the order of 10–100 Bq/g (bec-querels per gram), which precludes consi-dering it as conventional waste;– short-lived low and intermediate levelwaste (LILW-SL); radioactivity levels forsuch waste lie as a rule in a range from
a few hundred to one million Bq/g, ofwhich less than 10,000 Bq/g is from long-lived radionuclides. Its radioactivity beco-mes comparable to natural radioactivityin less than three hundred years.Production of such waste stands at some15,000 m3 per year in France;– long-lived low-level waste (LLW-LL);this category includes radium-bearingwaste from the extraction of rare earthsfrom radioactive ore, and graphite wastefrom first-generation reactors;– long-lived intermediate-level waste(ILW-LL), this being highly disparate, whe-ther in terms of origin or nature, with anoverall stock standing, in France, at45,000 m3 at the end of 2004. This mainlycomes from spent fuel assemblies (clad-ding hulls and end-caps), or from opera-tion and maintenance of installations; thisincludes, in particular, waste conditionedduring spent fuel reprocessing operations(as from 2002, this type of waste is com-pacted, amounting to some 200 m3
annually), technological waste from theoperation or routine maintenance of pro-duction or fuel-processing plants, fromnuclear reactors or from research cen-ters (some 230 m3 annually), along withsludges from effluent treatment (less than100 m3 annually). Most such waste gene-rates little heat, however some waste ofthis type is liable to release gases;– high-level waste (HLW), containing fis-sion products and minor actinides parti-tioned during spent fuel reprocessing (seeBox B), and incorporated at high tempe-rature into a glass matrix. Some 120 m3
of “nuclear glass” is thus cast every year.This type of waste bears the major partof radioactivity (over 95%), consequentlyit is the seat of considerable heat release,this remaining significant on a scale ofseveral centuries.Overall, radioactive waste conditioned inFrance amounts to less than 1 kg per year,per capita. That kilogram consists, forover 90%, of LILW-SL type waste, bearingbut 5% of total radioactivity; 9% of ILW-LL waste, less than 1% HLW, and virtuallyno LLW-LL waste.
What of the waste of tomorrow?From 1991, ANDRA compiled, on a yearlybasis, a geographical inventory of wastepresent on French territory. In 2001,ANDRA was asked by government to aug-ment this “National Inventory,” with thethreefold aim of characterizing extantstocks (state of conditioning, processing
traceability), predicting future waste pro-duction trends to 2020, and informing thepublic (see An inventory projecting into thefuture). ANDRA published this referenceNational Inventory at the end of 2004. Tomeet requirements for research in com-pliance with the directions set out in theFrench Act of 30 December 1991 (seeRadioactive waste management research:an ongoing process of advances), ANDRA,in collaboration with waste producers,has drawn up a Dimensioning InventoryModel (MID: Modèle d’inventaire dedimensionnement), for the purposes ofarriving at estimates of the volume ofwaste packages to be taken on board inresearch along direction 2 (disposal). Thismodel, including as it does predictions asto overall radioactive waste arisings fromthe current reactor fleet, over their entirelifespan, seeks to group waste types intofamilies, homogeneous in terms of cha-racteristics, and to formulate the mostplausible hypotheses, with respect toconditioning modes, to derive the volu-mes to be taken on board for the purpo-ses of the investigation. Finally, MID setsout to provide detailed stocktaking, inten-ded to cover waste in the broadest pos-sible fashion. MID (not to be confused withthe National Inventory, which has theremit to provide a detailed account ofactual waste currently present on Frenchterritory) thus makes it possible to bringdown the variety of package families to alimited number of representative objects,and to specify the requisite margins oferror, to ensure the design and assess-ment of disposal safety will be as robustas feasible, with respect to possible futurevariations in data.To ensure consistency between investi-gations carried out in accordance withdirection 2 and those along direction 3(conditioning and long-term storage), CEAadopted MID as input data. MID subsu-mes waste packages into standard pac-kage types, then computes the numberand volume of HLW and ILW-LL packa-ges, according to a number of scenarios,all based on the assumption that currentnuclear power plants will be operated for40 years, their output plateauing at400 TWhe per year.Table 1 shows the numbers and volumesfor each standard package type, for thescenario assuming a continuation of cur-rent strategy, with respect to spent fuelreprocessing: reprocessing of 79,200 UOXfuel assemblies and storage of 5,400 MOX
What is radioactive waste?
A
assemblies discharged from the currentPWR fleet, when operated over 40 years.
What forms does it come in?Five types of generic packages (also foundin MID) may be considered:• cementitious waste packages: ILW-LLwaste packages employing hydraulic-bin-der based materials as a conditioningmatrix, or as an immobilizing grout, or yetas a container constituent;• bituminized sludge packages: LLW andILW-LL waste packages, in which bitumenis used as confinement matrix for low- andintermediate-level residues from treat-ment of a variety of liquid effluents (fuelprocessing, research centers, etc.);• standard compacted waste packages(CSD-C: colis standard de déchets compac-tés): ILW-LL packages obtained throughcompaction conditioning of structural wastefrom fuel assemblies, and technologicalwaste from the La Hague workshops;• standard vitrified waste packages(CSD-V: colis standard de déchets vitrifiés):
HLW packages, obtained mainly throughvitrification of highly active solutions fromspent fuel reprocessing;• spent fuel packages: packages consis-ting in nuclear fuel assemblies dischargedfrom reactors; these are not considered tobe waste in France.The only long-lived waste packages to begenerated in any significant amounts bycurrent electricity production (see Box B)are vitrified waste packages and standardcompacted waste packages, the other typesof packages having, for the most part,already been produced, and bearing but asmall part of total radioactivity.
What is happening to this waste atpresent? What is to be done in thelong term?The goal of long-term radioactive wastemanagement is to protect humankind andits environment from the effects of thematerials comprised in this waste, mostimportantly from radiological hazards. Anyrelease or dissemination of radioactive
materials must thus be precluded, throughthe lasting isolation of such waste from theenvironment. This management is guidedby the following principles: to produce aslittle waste as practicable; limit its hazar-dous character as far as feasible; take intoaccount the specific characters of eachcategory of waste; and opt for measuresthat will minimize the burden (monitoring,maintenance) for future generations.As for all nuclear activities subject to controlby the French Nuclear Safety Authority(Autorité de sûreté nucléaire), fundamentalsafety regulations (RFSs: règles fonda-mentales de sûreté) have been drawn upwith respect to radioactive waste mana-gement: sorting, volume reduction, pac-kage confinement potential, manufactu-ring method, radionuclide concentration.RFS III-2.f, in particular, specifies the condi-tions to be met for the design of, anddemonstration of safety for an undergroundrepository, and thus provides a basic guidefor disposal investigations. Industrial solu-tions (see Industrial solutions for all low-level waste) are currently available for nighon 85% (by volume) of waste, i.e. VLLW andLILW-SL waste. A solution for LLW-LLwaste is the subject of ongoing investiga-tion by ANDRA, at the behest of waste pro-ducers. ILW-LL and HLW waste, contai-ning radionuclides having very longhalf-lives (in some cases, greater thanseveral hundred thousand years) are cur-rently held in storage installations comingunder the control of the Nuclear SafetyAuthority. What is to become of this wastein the long term, beyond this storage phase,is what the Act of 30 December 1991addresses (see Table 2).For all of these waste types, the FrenchNuclear Safety Authority is drawing up aNational Radioactive Waste ManagementPlan, specifying, for each type, a manage-ment pathway.
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 4,580waste packages
Total B 192,740 78,400Total overall 235,210 85,810
Short-lived Long-livedHalf-life < 30 years Half-life > 30 years
for the main elements
Very-low-level Morvilliers dedicated disposal facility (open since 2003)waste (VLLW) Capacity: 650,000 m3
Low-level waste Dedicated disposal facility under(LLW) investigation for radium-bearing
waste (volume: 100,000 m3)Aube Center and graphite waste
(open since 1992) (volume: 14,000 m3)Intermediate-level Capacity: 1 million m3
waste (ILW) MID volume estimate: 78,000 m3
High-level waste MID volume estimate: 7,400 m3
(HLW)
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’sDimensioning Inventory Model (MID).
Table 2. Long-term management modes, as currently operated, or planned, in France, byradioactive waste category. The orange area highlights those categories targeted byinvestigations covered by the Act of 30 December 1991.
(1) According to the Dimensioning Inventory Model (MID)
* renamed Areva NC in 2006
(next)
B
Most high-level (high-activity) radio-active waste (HLW) originates, in
France, in the irradiation, inside nuclearpower reactors, of fuel made up fromenriched uranium oxide (UOX) pellets, oralso, in part, from mixed uranium andplutonium oxide (MOX). Some 1,200 ton-nes of spent fuel is discharged annuallyfrom the fleet of 58 pressurized-waterreactors (PWRs) operated by EDF, sup-plying over 400 TWh per year, i.e. morethan three quarters of French nationalpower consumption.The fuel’s composition alters, during itsirradiation inside the reactor. Shortly afterdischarge, fuel elements contain, on ave-rage,(1) some 95% residual uranium, 1%plutonium and other transuranic ele-ments – up to 0.1% – and 4% of productsyielded by fission. The latter exhibit verysignificant radioactivity levels – to theextent this necessitates managementsafety measures requiring major indus-trial resources – of some 1017 Bq per tonneof initial uranium (tiU) (see Figure 1).The uranium found in spent fuel exhibitsa makeup that is obviously different fromthat of the initial fuel. The greater theirradiation, the higher the consumptionof fissile nuclei, and consequently thegreater the extent by which the uraniumwill have been depleted of the fissile iso-tope 235 (235U). Irradiation conditionsusually prevailing in reactors in the Frenchfleet, with an average fuel residence timeinside the reactor of some 4 years, for a
burnup rate close to 50 GWd/t, result inbringing down final 235U content to a valuequite close to that of natural uranium(less than 1%), entailing an energy poten-tial very close to the latter’s. Indeed, eventhough this uranium remains slightlyricher in the fissile isotope than naturaluranium, for which 235U content standsat 0.7%, the presence should also benoted, in smaller, though significant,amounts, of other isotopes having adverseeffects in neutronic or radiological terms(232U, 236U), that had not figured in theinitial fuel (see Table 1).
The plutonium present in spent fuel isyielded by successive neutron captureand decay processes. Part of the Pu isdissipated through fission: thus aboutone third of the energy generated is yiel-ded by “in situ recycling” of this element.These processes further bring about theformation of heavy nuclei, involving, whe-ther directly themselves, or through theirdaughter products, long radioactive half-lives. These are the elements of the acti-nide family, this including, essentially,plutonium (from 238Pu to 242Pu, the odd-numbered isotopes generated in partundergoing fission themselves duringirradiation), but equally neptunium (Np),americium (Am), and curium (Cm), knownas minor actinides (MAs), owing to the
Waste from the nuclear power cycle
(1) These figures should be taken as indicative values. They allow orders of magnitude to bepinpointed for enriched-uranium oxide fuel, taken from the main current French nuclear powerpathway; they do depend, however, on a number of parameters, such as initial fuel composition andirradiation conditions, particularly irradiation time.
Figure 1.The main elements found in spent nuclear fuel.
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 1,070U
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
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%)
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 areexpressed per tonne of initial uranium (tiU) for UOX, per tonne of initial heavy metal (tihm) for MOX.
H1
Li3
Na11
K19
Rb37
Cs55
Fr87
Be4
Mg12
Ca20
Sr38
Ba56
Ra88
Sc21
Y39
Ln
An
Ti22
Zr40
Hf72
Rf104
V23
Nb41
Ta73
Db105
Cr24
Mo42
W74
Sg106
Mn25
Tc43
Re75
Bh107
Fe26
Ru44
Os76
Hs108
Co27
Rh45
Ir77
Mt109
Ni28
Pd46
Pt78
Uun110
Ac89
Th90
Pa91
U92
Np93
Pu94
Am95
Cm96
Bk97
Cf98
Es99
Fm100
Md101
No102
Lr103
Cu29
Ag47
Au79
Zn30
Cd48
Hg80
Ga31
In49
TI81
Ge32
Sn50
Pb82
As33
Sb51
Bi83
Se34
Te52
Po84
Br35
I53
At85
Kr36
Al13
Si14
P15
S16
Cl17
Ar18
B5
C6
N7
O8
F9
Ne10
He2
Xe54
Rn86
La57
Ce58
Pr59
Nd60
Pm61
Sm62
Eu63
Gd64
Tb65
Dy66
Ho67
Er68
Tm69
Yb70
Lu71
■ heavy nuclei■ fission products
long-lived radionuclides
■ activation products■ fission and activation products
actinides
lanthanides
B (next)lesser abundance of these elements, com-pared with that of U and Pu, the latter beingtermed major actinides.Activation processes affecting nuclei of non-radioactive elements mainly involve struc-tural materials, i.e. the materials of thetubes, grids, plates and end-fittings thatensure the mechanical strength of nuclearfuel. These materials lead, in particular,to formation of carbon 14 (14C), with a half-life of 5,730 years, in amounts that arehowever very low, much less than one gramper tonne of initial uranium (g/tiU) in usualconditions.It is the products yielded by fission of theinitial uranium 235, but equally of the Pugenerated (isotopes 239 and 241), knownas fission products (FPs), that are theessential source of the radioactivity ofspent fuel, shortly after discharge. Over300 radionuclides – two thirds of whichhowever will be dissipated through radio-active decay in a few years, after irradia-tion – have been identified. These radio-nuclides are distributed over some40 elements in the periodic table, fromgermanium (32Ge) to dysprosium (66Dy),with a presence of tritium from fission, i.e.from the fission into three fragments (ter-nary fission) of 235U. They are thus cha-racterized by great diversity: diverse radio-active properties, involving as they do somehighly radioactive nuclides having very
short lifespans, and conversely othershaving radioactive half-lives counted inmillions of years; and diverse chemicalproperties, as is apparent from the ana-lysis, for the “reference” fuels used inPWRs in the French fleet, of the break-down of FPs generated, by families in theperiodic table (see Table 2). These FPs,along with the actinides generated, are,for the most part, present in the form ofoxides included in the initial uranium oxide,which remains by far the majority consti-tuent. Among some notable exceptionsmay be noted iodine (I), present in the formof cesium iodide, rare gases, such as kryp-ton (Kr) and xenon (Xe), or certain noblemetals, including ruthenium (Ru), rho-dium (Rh), and palladium (Pd), which mayform metallic inclusions within the oxidematrix.Pu is recycled nowadays in the form ofMOX fuel, used in part of the fleet (some20 reactors currently). Residual U may inturn be re-enriched (and recycled as a sub-stitute for mined uranium). Recyclingintensity depends on market prices fornatural uranium, the recent upturn inwhich should result in raising the currentrecycling rate (about one third being recy-cled at present).Such U and Pu recycling is the foundationfor the reprocessing strategy currentlyimplemented in France, for the major partof spent fuel (some two thirds currently).
For the 500 kg or so of U initially contai-ned in every fuel element, and after par-titioning of 475 kg of residual U and about5 kg Pu, this “ultimate” waste amountsto less than 20 kg of FPs, and less than500 grams MAs. This waste managementpathway (otherwise know as the closedcycle), consisting as it does in reproces-sing spent fuel now, to partition recove-rable materials and ultimate waste, dif-fers from strategies whereby spent fuelis conserved as-is, whether this be due toa wait-and-see policy (pending a decisionon a long-term management mode), or toa so-called open cycle policy, wherebyspent fuel is considered to be waste, anddesignated for conditioning into contai-ners, and disposal as-is.In the nuclear power cycle, as it is imple-mented in France, waste is subdivided intotwo categories, according to its origin.Waste directly obtained from spent fuel isfurther subdivided into minor actinidesand fission products, on the one hand, andstructural waste, comprising hulls (seg-ments of the cladding tubes that had heldthe fuel for PWRs) and end-caps (fittingsforming the end-pieces of the fuel assem-blies for these same PWRs), on the otherhand. The process used for spent fuelreprocessing, to extract U and Pu, alsogenerates technological waste (operatio-nal waste, such as spare parts, protec-tion gloves…) and liquid effluents.
UOX 33 GWd/tiU UOX 45 GWd/tiU UOX 60 GWd/tiU MOX 45 GWd/tihmfamily (E 235U: 3.5%) (E 235U: 3.7%) (E 235U: 4.5%) (Ei Pu: 8.65%)
quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tiU) quantity (kg/tihm)
rare gases (Kr, Xe) 5.6 7.7 10.3 7
alkali metals(Cs, Rb) 3 4 5.2 4.5
alkaline-earthmetals (Sr, Ba) 2.4 3.3 4.5 2.6
Y and lanthanides 10.2 13.8 18.3 12.4
zirconium 3.6 4.8 6.3 3.3
chalcogens(Se, Te) 0.5 0.7 1 0.8
molybdenum 3.3 4.5 6 4.1
halogens (I, Br) 0.2 0.3 0.4 0.4
technetium 0.8 1.1 1.4 1.1
Ru, Rh, Pd 3.9 5.7 7.7 8.3
miscellaneous:Ag, Cd, Sn, Sb… 0.1 0.2 0.3 0.6
Table 2.Breakdown by chemical family of fission products in spent UOX and MOX fuel, after 3 years’ cooling,for a variety of enrichment and burnup rates.
After discharge, spent fuel is stored in cooling pools, to allow its radioactivity to come down significantly. Shown here is a storage pool at Areva’s spent fuel reprocessing plant at La Hague.
Mag
num
/Har
ry G
ruya
ert
C
Raw, solid or liquid radioactive wasteundergoes, after characterization
(determination of its chemical and radio-logical makeup, and of its physical–che-mical properties), conditioning, a termcovering all the operations consisting inbringing this waste (or spent fuel assem-blies) to a form suitable for its transport,storage, and disposal (see Box D). Theaim is to put radioactive waste into asolid, physically and chemically stableform, and ensure effective, lasting confi-nement of the radionuclides it contains.For that purpose, two complementaryoperations are carried out. As a rule,waste is immobilized by a material –whether by encapsulation or homoge-neous incorporation (liquid or powderedwaste, sludges), or encasing (solid waste)– within a matrix, the nature of, and per-formance specification for which dependon waste type (cement for sludges, eva-poration concentrates and incinerationashes; bitumen for encapsulation ofsludges or evaporation concentratesfrom liquid effluent treatment; or avitreous matrix, intimately binding thenuclides to the glass network, for fis-sion product or minor actinide solutions).This matrix contributes to the confine-ment function. The waste thus conditio-ned is placed in an impervious contai-
ner (cylindrical or rectangular), consis-ting in one or more canisters. The whole– container and content – is termed apackage. Equally, waste may be com-pacted and mechanically immobilizedwithin a canister, the whole forming apackage.When in the state they come in as sup-plied by industrial production, they areknown as primary packages, the pri-
mary container being the cement ormetal container into which the conditio-ned waste is ultimately placed, to allowhandling. The container may act as initialconfinement barrier, allotment of func-tions between matrix and container beingdetermined according to the nature ofthe waste involved. Thus, the whole obtai-ned by the grouping together, within onecontainer, of a number of primary
Cross-section of an experimental storage borehole for a spent fuel container (the lower part of theassembly may be seen, top right), in the Galatée gallery of CECER (Centre d’expertise sur leconditionnement et l’entreposage des matières radioactives: Radioactive Materials Conditioningand Storage Expertise Center), at CEA’s Marcoule Center, showing the nested canisters.
A. G
onin
/CEA
What stands between waste and the environment?
C (next)ILW-LL packages may ensure confinementof the radioactivity of this type of waste.If a long-term storage stage is found tobe necessary, beyond the stage of indus-trial storage on the premises of the pro-ducers, primary waste packages mustbe amenable to retrieval, as and whenrequired: durable primary containersmust then be available, in such condi-tions, for all types of waste.In such a case, for spent fuel assemblieswhich might at some time be earmar-ked for such long-term storage, or evenfor disposal, it is not feasible to demons-trate, on a timescale of centuries, theintegrity of the cladding holding the fuel,forming the initial confinement barrierduring the in-reactor use stage. Securingthese assemblies in individual, imper-vious cartridges is thus being conside-red, this stainless-steel cartridge beingcompatible with the various possiblefuture management stages: treatment,return to storage, or disposal. Placingthese cartridges inside imperviouscontainers ensures a second confine-ment barrier, as is the case for high-level waste packages.In storage or disposal conditions, thewaste packages will be subjected to avariety of aggressive agents, both inter-nal and external. First, radionuclide
radioactive decay persists inside the pac-kage (self-irradiation process). Emissionof radiation is concomitant with heatgeneration. For example, in confinementglasses holding high-activity (high-level)waste, the main sources of irradiationoriginate in the alpha decay processesfrom minor actinides, beta decay fromfission products, and gamma transitions.Alpha decay, characterized by produc-tion of a recoil nucleus, and emission ofa particle, which, at the end of its path,yields a helium atom, causes the majorpart of atom displacements. In particu-lar, recoil nuclei, shedding considerableenergy as they do over a short distance,result in atom displacement cascades,thus breaking large numbers of chemi-cal bonds. This is thus the main causeof potential long-term damage. In suchconditions, matrices must exhibit ther-mal stability, and irradiation-damageresistance.Stored waste packages will also be sub-jected to the effects of water (leaching).Container canisters may exhibit a deg-ree of resistance to corrosion processes(the overpacks contemplated for glas-ses may thus delay by some 4,000 yearsthe arrival of water), and the confine-ment matrices must be proven to exhi-bit high chemical stability.
Between the containers and the ultimatebarrier provided, in a radioactive wastedeep disposal facility, by the geologicalenvironment itself, there may further beinterposed, apart, possibly, from an over-pack, other barriers, so-called engi-neered barriers, for backfill and sealingpurposes. While these would be point-less as backfill in clay formations, theywould have the capability, in other envi-ronments (granite), of further retardingany flow of radionuclides to the geo-sphere, notwithstanding degradation ofthe previously mentioned barriers.
CEA
Technological demonstratorsof ILW-LL packages for
bituminized sludges.
D
The object of nuclear waste storageand disposal is to ensure the long-
term confinement of radioactivity, inother words to contain radionuclideswithin a definite space, segre-gated from humankind and theenvironment, as long as requi-red, so that the possible returnto the biosphere of minuteamounts of radionuclides canhave no unacceptable health orenvironmental impact.According to the Joint Con-vention on the Safety of SpentFuel Management and on theSafety of Radioactive WasteManagement, signed on 5 Sep-tember 1997, “storage” means“the holding of spent fuel or ofradioactive waste in a facility thatprovides for its containment,with the intention of retrieval.”This is thus, by definition, aninterim stage, amounting to adelaying, or wait-and-see solu-tion, even though this may be fora very long time (from a fewdecades to several hundredyears), whereas disposal may befinal.Used from the outset of the nuclearpower age, industrial storage keepsspent fuel awaiting reprocessing, andconditioned high-level waste (HLW), orlong-lived intermediate-level waste
(ILW-LL) in conditions of safety, pen-ding a long-term management modefor such waste. Retrieval of stored pac-kages is anticipated, after a period oflimited duration (i.e. after a matter of
years, or tens of years).Long-term storage (LTS) may becontemplated, in particular, in the eventof the deferred deployment of a dispo-sal facility, or of reactors to carry out
recycling–transmutation, or simply toturn to advantage the natural decay ofradioactivity (and hence the falling offof heat release from high-level waste),before putting the waste into geologi-
cal disposal. By “long term” ismeant a timespan of up to 300years. Long-term storage maytake place in a surface or sub-surface facility. In the formercase, the site may be protected,for instance, by a reinforced-concrete structure. In the lattercase, it will be located at a depthof some tens of meters, and pro-tected by a natural environment(for instance, if buried in a hill-side) and its host rock.Whichever management stra-tegy is chosen, it will be impe-rative to protect the biospherefrom the residual ultimate waste.The nature of the radioelementsthe latter contains means a solu-tion is required that has the abi-lity to ensure their confinementover several tens of thousandyears, in the case of long-livedwaste, or even longer. On suchtimescales, social stability is amajor uncertainty that has to be
taken on board. Which is why disposalin deep geological strata (typically, 500m down) is seen as a reference solu-tion, insofar as it inherently makes fordeployment of a more passive techni-cal solution, with the ability to stand,with no increased risk, an absence ofsurveillance, thus mitigating a possibleloss of memory on the part of society.The geological environment of such adisposal facility thus forms a further,essential barrier, which does not existin the storage case.A disposal facility may be designed tobe reversible over a given period. Theconcept of reversibility means the designmust guarantee the ability, for a varietyof reasons, to access the packages, oreven to take them out of the facility, overa certain timespan, or to opt for the finalclosure of the disposal facility. Suchreversibility may be envisaged as a suc-cession of stages, each affording adecreasing “level of reversibility.” Tosimplify, each stage consists in carryingout one further technical operation brin-ging the facility closer to final closure,making retrieval more difficult than atthe previous stage, according to well-specified criteria.
From storage to disposal
ANDRA design for the disposal of standard vitrified waste packages in horizontal galleries,showing in particular the packages’ various canisters, and some characteristics linked to potential reversibility of the disposal facility.
CEA design study for a common container for the long-term storage and disposal of long-lived, intermediate-level waste.
AN
DR
A
A. G
onin
/CEA
lid
1.30m to 1.60 m
0.57 m to 0.64 m
primary package
disposal packageslining
intercalary
excavated diameter: 0.7 m approx.
length: 40 m approx.
steel overpack
E
Transmutation is the transformationof one nucleus into another, through
a reaction induced by particles with whichit is bombarded. As applied to the treat-ment of nuclear waste, this consists inusing that type of reaction to transformlong-lived radioactive isotopes into iso-topes having a markedly shorter life, oreven into stable isotopes, in order toreduce the long-term radiotoxic inven-tory. In theory, the projectiles used maybe photons, protons, or neutrons.In the first case, the aim is to obtain, bybremsstrahlung,(1) through bombardmentof a target by a beam of electrons, pro-vided by an accelerator, photons able tobring about reactions of the (γ, xn) type.Under the effects of the incoming gammaradiation, x neutrons are expelled fromthe nucleus. When applied to substan-ces that are too rich in neutrons, andhence unstable, such as certain fissionproducts (strontium 90, cesium 137…),such reactions yield, as a rule, stable sub-stances. However, owing to the very lowefficiency achieved, and the very highelectron current intensity required, thispath is not deemed to be viable.In the second case, the proton–nucleusinteraction induces a complex reaction,known as spallation, resulting in frag-mentation of the nucleus, and the release
of a number of particles, including high-energy neutrons. Transmutation by wayof direct interaction between protons isuneconomic, since this would involve, inorder to overcome the Coulomb barrier,(2)
very-high-energy protons (1–2 GeV),requiring a generating energy greaterthan had been obtained from the processthat resulted in producing the waste. Onthe other hand, indirect transmutation,using very-high-energy neutrons (ofwhich around 30 may be yielded, depen-ding on target nature and incoming protonenergy), makes it possible to achieve verysignificantly improved performance. Thisis the path forming the basis for thedesign of so-called hybrid reactors, cou-pling a subcritical core and a high-inten-sity proton accelerator (see Box F, Whatis an ADS?).The third particle that may be used is thusthe neutron. Owing to its lack of electriccharge, this is by far the particle best sui-ted to meet the desired criteria. It is “natu-rally” available in large quantities insidenuclear reactors, where it is used to trig-ger fission reactions, thus yielding energy,while constantly inducing, concurrently,transmutations, most of them unsought.The best recycling path for waste wouldthus be to reinject it in the very installa-tion, more or less, that had produced it…
When a neutron collides with a nucleus,it may bounce off the nucleus, or pene-trate it. In the latter case, the nucleus,by absorbing the neutron, gains excessenergy, which it then releases in variousways:• by expelling particles (a neutron, e.g.),while possibly releasing radiation;• by solely emitting radiation; this isknown as a capture reaction, since theneutron remains captive inside thenucleus;• by breaking up into two nuclei, of moreor less equal size, while releasing concur-rently two or three neutrons; this is knownas a fission reaction, in which considera-ble amounts of energy are released.Transmutation of a radionuclide may beachieved either through neutron captureor by fission. Minor actinides, as elementshaving large nuclei (heavy nuclei), mayundergo both fission and capture reac-tions. By fission, they transform intoradionuclides that, in a majority of cases,are short-lived, or even into stable nuclei.The nuclei yielded by fission (known asfission products), being smaller, are onlythe seat of capture reactions, undergoing,on average, 4 radioactive decays, with ahalf-life not longer than a few years, asa rule, before they reach a stable form.Through capture, the same heavy nucleitransform into other radionuclides, oftenlong-lived, which transform in turnthrough natural decay, but equallythrough capture and fission.
What is transmutation?
(1) From the German for “braking radiation.” High-energy photon radiation, yielded by accelerated(or decelerated) particles (electrons) following a circular path, at the same time emitting brakingphotons tangentially, those with the highest energies being emitted preferentially along the electronbeam axis.
(2) A force of repulsion, which resists the drawing together of same-sign electric charges.
E (next)
The probability, for a neutron, of causinga capture or a fission reaction is evalua-ted on the basis, respectively, of its cap-ture cross-section and fission cross-sec-tion. Such cross-sections depend on thenature of the nucleus (they vary consi-derably from one nucleus to the next, and,even more markedly, from one isotopeto the next for the same nucleus) andneutron energy.For a neutron having an energy lowerthan 1 eV (in the range of slow, or ther-mal, neutrons), the capture cross-sec-
tion prevails; capture is about 100 timesmore probable than fission. This remainsthe case for energies in the 1 eV–1 MeVrange (i.e., that of epithermal neutrons,where captures or fissions occur at defi-nite energy levels). Beyond 1 MeV (fastneutron range), fissions become moreprobable than captures.Two reactor pathways may be conside-red, according to the neutron energyrange for which the majority of fissionreactions occur: thermal-neutron reac-tors, and fast-neutron reactors. The ther-
mal neutron pathway is the technologyused by France for its power generationequipment, with close to 60 pressurized-water reactors. In a thermal-neutronreactor, neutrons yielded by fission areslowed down (moderated) through colli-sions against light nuclei, making upmaterials known as moderators. Due tothe moderator (common water, in thecase of pressurized-water reactors), neu-tron velocity falls off, down to a few kilo-meters per second, a value at which neu-trons find themselves in thermalequilibrium with the ambient environ-ment. Since fission cross-sections for235U and 239Pu, for fission induced bythermal neutrons, are very large, aconcentration of a few per cent of thesefissile nuclei is sufficient to sustain thecascade of fissions. The flux, in a ther-mal-neutron reactor, is of the order of1018 neutrons per square meter, persecond.In a fast-neutron reactor, such as Phénix,neutrons yielded by fission immediatelyinduce, without first being slowed down,further fissions. There is no moderator inthis case. Since, for this energy range,cross-sections are small, a fuel rich infissile radionuclides must be used (up to20% uranium 235 or plutonium 239), if theneutron multiplication factor is to be equalto 1. The flux in a fast-neutron reactor isten times larger (of the order of 1019 neu-trons per square meter, per second) thanfor a thermal-neutron reactor.
Figure.Simplified representation of the evolution chain of americium 241 in a thermal-neutron reactor(shown in blue: radionuclides disappearing through fission). Through capture, 241Am transformsinto 242mAm, this disappearing predominantly through fission, and into 242Am, which mainly decays(with a half-life of 16 hours) through beta decay into 242Cm. 242Cm transforms through alpha decayinto 238Pu, and through capture into 243Cm, which itself disappears predominantly through fission.238Pu transforms through capture into 239Pu, which disappears predominantly through fission.
241Am432ans
242Cm163days
242Am16h
242mAm141
years
238Pu87.8
years
240Pu6.5.103
years
239Pu2.4.104
years
241Pu14.4
years
243Cm32
years
86%
98%36%82%
capture
fission
alphadecay
beta-minusdecay
electroncapture
24%
86%
4%
1%95%
11%
15%
12%
2%
5%13% 64% 2% 69%
85%14%
7%
3%86%
F
An ADS (accelerator-driven system) isa hybrid system, comprising a
nuclear reactor operating in subcriticalmode, i.e. a reactor unable by itself to sus-tain a fission chain reaction, “driven” byan external source, having the ability tosupply it with the required comple-ment of neutrons.(1)
Inside the core of a nuclear reactor,indeed, it is the fission energy fromheavy nuclei, such as uranium 235or plutonium 239, that is released.Uranium 235 yields, when under-going fission, on average 2.5 neu-trons, which can in turn induce afurther fission, if they collide with auranium 235 nucleus. It may thusbe seen that, once the initial fissionis initiated, a chain reaction may develop,resulting, through a succession of fis-sions, in a rise in the neutron population.However, of the 2.5 neutrons yielded bythe initial fission, some are captured, thusnot giving rise to further fissions. Thenumber of fissions generated from oneinitial fission is characterized by the effec-tive multiplication factor keff, equal to theratio of the number of fission neutronsgenerated, over the number of neutronsdisappearing. It is on the value of this coef-ficient that the evolution of the neutronpopulation depends: if keff is markedlyhigher than 1, the population increasesrapidly; if it is slightly higher than 1, neu-tron multiplication sets in, but remainsunder control; this is the state desired atreactor startup; if keff is equal to 1, thepopulation remains stable; this is the state
for a reactor in normal operating condi-tions; and, if keff is lower than 1, the neu-tron population dwindles, and becomesextinct, unless – as is the case for a hybridsystem – an external source provides aneutron supply.
From the effective multiplication factor,a reactor’s reactivity is defined by the ratio(keff –1)/keff. The condition for stability isthen expressed by zero reactivity. To sta-bilize a neutron population, it is sufficientto act on the proportion of materials exhi-biting a large neutron capture cross-sec-tion (neutron absorber materials) insidethe reactor.In an ADS, the source of extra neutronsis fed with protons, generated with anenergy of about 100 keV, then injectedinto an accelerator (linear accelerator orcyclotron), which brings them to an energyof around 1 GeV, and directs them to aheavy-metal target (lead, lead–bismuth,tungsten or tantalum). When irradiatedby the proton beam, this target yields,through spallation reactions, an intense,high-energy (1–20 MeV) neutron flux, onesingle incoming neutron having the abi-lity to generate up to 30 neutrons. The lat-
ter then go on to interact with the fuel ofthe subcritical neutron multipliermedium, yielding further neutrons (fis-sion neutrons) (see Figure).Most hybrid system projects use as a core(of annular configuration, as a rule) fast-
neutron environments, since thesemake it possible to achieve neu-tron balances most favorable totransmutation, an operation thatallows waste to be “burned,” butwhich may equally be used to yieldfurther fissile nuclei. Such a sys-tem may also be used for energygeneration, even though part ofthis energy must be set aside topower the proton accelerator, apart that is all the higher, the more
subcritical the system is. Such a systemis safe in principle from most reactivityaccidents, its multiplication factor beinglower than 1, contrary to that of a reac-tor operated in critical mode: the chainreaction would come to a halt, if it wasnot sustained by this supply of externalneutrons.A major component in a hybrid reactor,the window, positioned at the end of thebeam line, isolates the accelerator fromthe target, and makes it possible to keepthe accelerator in a vacuum. Traversedas it is by the proton beam, it is a sensi-tive part of the system: its lifespandepends on thermal and mechanicalstresses, and corrosion. Projects are moo-ted, however, of windowless ADSs. In thelatter case, it is the confinement cons-traints, and those of radioactive spalla-tion product extraction, that must be takenon board.
What is an ADS?
(1) On this topic, see Clefs CEA, No. 37, p. 14
Principle schematic of an ADS.
window
protonsource
accelerator
subcritical reactor
spallationtarget
providingexternalneutrons
100 keV 1 GeV
The characteristics of the major part of the radioactive waste generated in France are determined by those of the French nuclearpower generation fleet, and of the spent fuel reprocessing plants, built in compliance with the principle of reprocessing such fuel, topartition such materials as remain recoverable for energy purposes (uranium and plutonium), and waste (fission products and minoractinides), 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 theMelox 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 pres-surized-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 opera-ted at La Hague (Manche département), since 1994, by Areva NC (formerly Cogema,) another member of the Areva Group (the UP3plant, put on stream in 1990–92, for its part, carries out reprocessing of fuel from other countries). The waste vitrification workshopsat these plants, the outcome of development work initiated at Marcoule, give their name (R7T7) to the “nuclear” glass used for theconfinement 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. Thesecould be fast-neutron reactors (i.e. fast reactors [FRs]), either sodium-cooled (SFRs) or gas-cooled (GFRs). Following the closingdown of the Superphénix reactor, in 1998, only one FR is operated in France, the Phénix reactor, due to be closed down in 2009.
The industrial context 1
(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.