Nuclear energyEnvironment, sustainable development,

Is it a renewable energy?

Robert Guillaumont

Part 1- Understanding nuclear energy

Part 2 - Considerations on nuclear energy

With regard to technological aspects the implementation of nuclear energy does differ much from others energies. Nuclear reactors and nuclear power plants (NPP) are large industrial facilities which concentrate and use innovative engineered components. What makes the difference is that NPP concentrate the biggest man-made radioactivity produced on earth.

This lecture will not focus on technological aspects but will focus on the points which make the difference and are sometimes not well understood in debates. In other words it will focus on the specificities of nuclear energy.

Very few quantitative comparisons will be made with other energies. They can be found everywhere in any consideration on energies.

Nuclear energy

The main problem to make nuclear energy safe is to avoid uncontrolled dissemination of radioactive matter.

To discuss the topics of i) nuclear energy and environment, ii) nuclear energy and sustainable development and iii) nuclear energy and renewable energy, the following information focus on nuclear reactors and on the radioactive matter handed in the nuclear fuel cycle, which could raise problems, from mine to radwastes.

Only few points are presented. Full information is given in the accompanying paper.

Nuclear reactor

A nuclear reactor allows to control the release of nuclear energy, which comes from some nuclear events: spontaneous decay (, , sf, ..) and induced neutron nuclear reactions: (n,f), (n,), (n,2n). Nuclides, particles and photons are produced with high energy (MeV to 100 MeV). Their kinetic or electromagnetic energy is lost into matter where nuclear processes occur (nuclear fuel), giving heat, later transformed in work according to thermodynamic laws. A nuclear electric reactor give electricity on the grid (0.5 to 1.3 GWe).

The fresh fuel contains special heavy nuclides (U or U-Pu isotopes) and later many others nuclides (1/3 of the chemical elements). The neutrons are produced in situ.

To understand the main characteristics of nuclear energy some basic definitions are mandatory.

The probability that a nuclide AZX disappears at t, during dt, is

  dPt, t+dt(1) = ( + )dt = dt,

(t-1) is the radioactive constant of a spontaneous decay event, (b) is the cross section of a nuclear process in the incoming flux of given particles on A

ZX ( (particles cm-2t-1). 1 barn = 10-24 cm2

When a nuclide, i, disappears a new nuclide (for , -, (n, ), (n,2n), or several new nuclides (for sf, (n,f), ..) appear.

For a given number N of nuclides one can shows that N = N0 exp-t

For spontaneous decay ( = ) the half-life T= 0.7/ is defined by N = N0/2. The activity of N radionuclides is A = N (Bq),

1Bq = 1 decay s-1, 1 Bq = 27 picoCurie, 1 Ci = 3.7 1010 Bq = 37 Gbq

For a nuclear reaction on a stable nuclide ( = 0) or a long-lived nuclide ( close to zero) the apparent half-life is T* = 0.7/ (t).

Otherwise the situation is very complicated and the build-up kinetics of nuclides disappearing and appearing by decays and nuclear reactions can be calculated providing , and values are known. Today coherent and updated international Data bank exist as well as softwares.

For any nuclear reaction on light nuclides leading to the capture of a particle is noted a

For (n,) reaction on heavy nuclides, is noted c

For (n,f) reaction on heavy nuclides, is noted f

f values depend on En (kT to few MeV)When En < 1 eV a neutron is slow, when En > 1 MeV a neutron is fast.When, for all En values, f > c a heavy nuclide is fissile, when c > f a heavy nuclide is fertile

Neutrons loss their energy by elastic collisions with nucleus (or disappear by nuclear process). Lower is A, higher is the energy loss (20 collisions for H, 25 for D, 120 for C, 2000 for U). Neutron born with few MeV of energy have a “range” before to be in thermal equilibrium (2.8 cm for H2O, 5.2 cm for C, 10 cm for D2O)

Table 1. Cross section (barn = 10-24 cm2) of “neutron absorption” for some nuclear processes in relation with elements present in a reactor (moderator, stainless steel, control rods, spent fuel).

Nuclide Reaction (n,) Reaction (n,p) Reaction (n)H1 0.330 (gives D)Li6 940 (gives T)B10 3840 (gives Li7)N14 1.80 (gives C14) O17 0.240 (gives C14)Cl35 43,6Mo95, 97 14.5, 2.5Ag107, 109 37, 90In115 115Xe131, 135 (T = 9h) 90, 2.85 106

Cs133 28Pr141 11Nd142, 143, 145 19, 330, 47Sm149,150,152 40.1104, 100, 200Eu151,153 9200, 370Gd155, 157 6 103, 2.54105

Cr50, 53 16, 18Mn55 13Fe, all isotopes 2.5Co59 30Ni58, 62 4.5, 95

The main phenomena which occur in nuclear fuel are i) the fission of fissile nuclides giving neutrons, ii) the build up of actinides from the fertile nuclides, iii) activation of materials. All are linked to the fate of neutrons (decrease of their energy, captures by nuclides, escapes).

All the phenomena give large amount of nuclear energy. One speaks, comparing to classical fuel, on the “burning of nuclear fuel”. But nuclear fuel behaves from a thermodynamic point of view as a closed system and not as an open system (only heat is exchanged). Neutrons play the role of oxygen.

Fission

Fission occur mainly with slow neutrons on U235 and Pu239, Pu241.. (A odd, Z even nuclides in general, but exceptions) and with fast neutrons on U238, Pu240, ..(A even, Z even nuclides in general). Fission yields depend on excitation function f = f(En). f decreases as 1/v for thermal neutrons (kT = 1/2 mnv2 =1/40 eV at 300 K), is stable for En> 1 MeV and shows “resonances” when En is in the range of keV (epithermal neutrons).

Fission of heavy nuclides gives fission fragments (FF), prompt rays (emitted by excited FF) and fast neutrons ( = 2 to 3) in a very short time (10-15 to 10-13 s for neutrons and 10-11 to 10-8 s for rays)

Around 90 % of FF are stable nuclides and 10 % are radionuclides. Each radioactive FF has an excess of neutrons and decay to a final stable nuclide through 5 to 6 - decays, giving daughters. FF and these daughters are called fission products (FP).

In some decay chains there are nuclides in high excited state (up to 1.5%), born by bêta decay, which decay by emission of fast neutron. These neutrons are called delayed neutron (1 to 80 s) compared to the prompt neutrons. There are the key for operating a reactor (Br87 for instance)

FP are artificial elements: 30 < Z < 66, 76 < A <161, with A centred around 97 (light FP) and 137 (heavy FP).

Half-lives of FP (- emitters) spread over milliseconds to millions of years.

75 100 125 150

0.001

0.01

0.1

0

1

10 U235

+ 0.1 % T

97 137

118

%

Sr90

Zr95Tc99

Ru106 I129

Xe135Cs137

Ce144 U235 Pu239

Fast neutrons

Thermal neutrons

Induced fission with thermal neutrons

A

Kr85

U235

Log10

Log10 En(eV)0

1

2

3

4

43210- 1- 2- 3

U235 Pu239

resonances

Fast neutrons

Slow neutrons

Table 3. Characteristic of delayed neutron for the fission of U235 induced by thermal neutrons. Variation of the fraction of delayed neutrons, (pcm = % 10-5) for the fission of other nuclides

Nuclide T(s) (s) N/fission (pcm) En(MeV)10-3

Br87 55.7 80.4 5.2 21 0.250I137,Br88 22.72 32.8 3.46 142I138,Br89 6.2 8.9 3.10 123I129, Br90 2.3 3.3 6.24 257I140 0.6 0.88 1.82 75Total 18 650

Th232 4000U233 300U235 650U238 800Pu239 200Am241 30Cm242 40

Short-lived FP have half-life less than 30 years

Half life of long-lived FP

Se79 6.3 104 y

Zr93 1.5106 y

Tc99 2.1 106 y

Pd107 6.5 106 y

I129 1.7 107 y

Cs135 3 106 y

Fate of neutrons

Fast neutrons born in fuel loss their kinetic energy by elastic collisions, induce fissions of U235 (and the process can propagate itself) and give also nuclear reactions. But as f (fast neutron) < f (slow neutron) the better yield of fission is obtained with thermal neutron. Fast neutron are slowed down with a moderator. When En reach kT the fuel is in a bath of “gaseous neutrons”. In this case (n cm2s-1) can be expressed as (n cm2v) with v = (2kT/mn)1/2 it is to say as (n cm3).

Build-up of actinides

Actinides (Z > 98, emitters except Pu241) are produced through nuclear reactions and decays in complicated ways.

(n,) :U235(n,)U236(n,) followed by bêta decays :U238(n,,2-)Pu239(n,2n) :Pu239(n,2n)Pu238(n,2n) followed by bêta decay:U238(n,2n,-)Np237(n,2n,-)Pu236. (,n) on light nuclides, if present.

Table 2. Cross section (barn = 10-24 cm2) for neutron induced fission reactions on some actinides (average values over neutrons energies)

PWR UOX, PWR MOX FNR UOX MOX, thermal n epithermal n fast nf c f c f c Z,N

U235 38.8 8.7 12.6 4.2 1.98 0.57 p,iU238 0.103 0.86 0.124 0.8 0.04 0.30Pu238 2.4 27.7 1.9 8 1.1 0.58 p,iPu239 102 58.7 21.7 12 1.86 0.56Pu240 0.53 210.2 0.7 24.6 0.36 0.57Pu241 102.2 40.9 28.5 9 2.49 0.47 p,iPu242 0.44 28.8 0.5 12.3 0.24 0.44Np237 0.52 33 0.6 18 0.32 1.7Am241 1.1 110 0.8 35.6 0.27 2,0 i,pAm242 159 301 3.2 0.6 i,iAm242m595 137 126.6 27.5 3.3 0.6 i,iAm243 0.44 49 0.5 31.7 0.21 1.8Cm242 1.14 4.5 0.96 3.45 0.58 1.0Cm243 88 14 43.1 7.32 7.2 1.0 p,iCm244 1.0 16 1 13.1 0.42 0.6Cm245 116 17 33.9 5.4 5.1 0.9 p,i

Building of actinides

234 235 236 237 238 239 240 241 242 243 244 245 246

Cm

Am

Pu

Np

U

75%

75%

71%

85%

m

n,2nn,

EC n,f

Main road

Half-life of actinides (in year)

U234 2.45 105 , U235 7.08 108, U236 2.34 107 U238 4.49 109

Pu238 87.7, Pu239 2.41104, Pu240 6.56103, Pu241 14.4, Pu24 3.7105

Np237 2.14 106

Am241 432.2, Am242m 152, Am243 7.38 103

Cm243 28.5, Cm244 18.1 Cm245 8.5 103

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18s d p

1 H1

He2

2 Li3

Be4

B5

C6

N7

O8

F9

Ne10

3 Na11

Mg12

Al13

Si14

P15

S16

Cl17

Ar18

4 K19

Ca20

Sc21

Ti22

V23

Cr24

Mn25

Fe26

Co27

Ni28

Cu29

Zn30

Ga31

Ge32

As33

Se34

Br35

Kr36

5 Rb37

Sr38

Y39

Zr40

Nb41

Mo42

Tc43

Ru44

Rh45

Pd46

Ag47

Cd48

In49

Sn50

Sb51

Te52

I53

Xe54

6 Cs55

Ba56

La*57

Hf72

Ta73

W74

Re75

Os76

Ir77

Pt78

Au79

Hg80

Tl81

Pb82

Bi83

Po84

At85

Rn86

7 Fr87

Ra88

Ac**89

Rf104

Db105

Sg106

Bh107

Hs108

Mt109 110 111 112 113 114 115

**5f

Th90

Pa91

U92

Np93

Pu94

Am95

Cm96

Bk97

Cf98

Es98

Fm100

Md101

No102

Lr103

*4f

Ce58

Pr59

Nd60

Pm61

Sm62

Eu63

Gd64

Tb65

Dy66

Ho67

Er68

Tm69

Yb70

Lu71

Categorisation of nuclear Reactors

Thermal neutrons based (large majority)

They are complex assemblies of 3 main materials: nuclear fuel, moderator and coolant. The fuel contains fissile/fertile nuclides, the moderator slow down the fast neutrons from 2 MeV to kT and the coolant extract the heat produced in the fuel. The right number of neutron is adjusted with control material (a adapted)

Fast neutrons based (few)

They do not need moderator

Fuel

The natural radioelements in large amount which contains fissile/fertile nuclides are U (Unat, U235 0.71 %, U238, 99.3 %) and mono-isotopic Th (Th232). U235 is the only one which has f > c whatever En is (fissile nuclide). U238 and Th232 are fertile nuclides because the processes (n,,-) give Pu239 and U233, which are fissile nuclides.

Nuclear energy lies, up to now on Unat (see later for Th).

An artificial radioelement which contains fissile nuclides, is Pu (Pu238 to 242, mainly Pu239, Pu240 and Pu241). When available as separated element (it is produced in uranium irradiated fuel) it is used associated with U as fuel.

Moderators - low Z, low a : H2O, D2O, C (as graphite)Coolants - low Z, low a :H2O, D2O, CO2, He.Control rods - moderate or high a : B, Ag, Gd Unat (metal) can be used with C and CO2 (GGR, Magnox)Unat can be used with D2O (CANDU). D2O needs electrical energyUenr (enriched in U235, 3 to 5 %) can be associated with H2O (PWR and BWR). 1 ton of Uenr between 2.3 to 4 % need 4 to 9 tons of Unat, assuming tails at 0.25 to 0.3 % for depleted U (Udep). The need of electrical energy is between 7.2 to 12 GWhe when enrichment is made by gaseous diffusion.

Control of fission propagation

The fission reaction can be propagated if one neutron issued from one fission can give one additional fission, the (-1) other can be lost.

In a reactor the multiplication factor keff = F (i ni fi/ ni a

i) must remains always equal to one or the reactivity = keff - 1 must remains always equal to zero. a is the cross section of any process which lead to the disappearance of neutron, ni is the number of any nuclide per unit of volume and summations is extended to all the different nuclides of the medium, F take into account losses of neutrons, moderation ratio, other parameters...

The time, , between two neutrons generations, g and g+1, is very short, some milliseconds for thermal neutrons and some microseconds for fast neutrons. P = P0 exp (keff-1)t/ (P = exp 10 in 1 s, = 0.1ms, keff = 1.001). Impossible to control the propagation of neutrons!

Fortunately there are % of delayed neutron (coming as said from some FP which are emitted seconds after the prompt neutrons of fission.

A reactor runs with keff < 1, (p < 0), for prompt neutrons and with keff = 1, (tot = 0) for the total neutrons, prompt and delayed (tot = p + d). In other words the number of neutron is adjusted with delayed neutrons to maintain keff = 1 or = 0.

Higher is, easier is the control of the propagation of fission. It decreases when A and Z increase and the control of the reactor is more difficult.

An important factor is T. When T change many parameters linked to keff change (F factor) and En change it is to say f and c. Drastic effects are for En >10 eV in resonances range (En = 0,1 eV). k/k is < 0 (U235) or > 0 (Pu239). Negative value equals passive safety.

As long as fission and nuclear reactions occur the FP and actinides accumulate in the nuclear fuel and the conditions for keeping keff = 1 are finally not longer possible. The fuel has to be renewed. The unloaded fuel is called spent fuel (SF).

Energy released

The total nuclear energy “associated” to one fission is, in average, around 200 MeV, counting the neutrons energies En (breakdown for a given nuclide between FF, other nuclear reactions, decays and losses is complicated)

According to : 1 eV.molecule = 23.0609 kcal.mole = 96.5098 kJ.mole and 1J = 0.2778 kWh, 1g of a given fissile nuclide with A = 235 (for instance) gives 34.3 107 kcal or 8.00 1010 J or 22 780 kWh (or 1.8 tep), which is over and over the energy given by the combustion of 1 g of any material.

For a given fissile/fertile matter the exact value depend on the isotopic composition of the fissile nuclide.

The « Burn Up » (BU) of a nuclear material is the energy (MWj) given by unit of weight (t) of IHM (initial heavy metal, no distinction between the fissile nuclides).

1 MWjt-1 correspond exactly to the release of 8.64 1010 J by ton of fissile and fertile nuclides. This value is very close to the energy given by 1 g of fissile material (1.053 with the scoping value based on A = 235).

1 MWjt-1 = 1.053 g of fissile nuclide = 1.053 g of FP (and actinides).

1 MW electric power (MWe) requires 3 MW thermal power (3 MWth) which needs each second the fission of 0.0365 mg of fissile nuclides.

In a material burnt to 45 GWjt-1 about 47 kg of fissile nuclides have been transformed into FP.

The electric power of a typical modern nuclear reactor is 1 GWe. It «burns» each second 36.5 mg of fissile nuclides. If it works 310 days a year (loading factor 85 %), 977 kg of fissile nuclides have disappeared giving the same amount of FP (and other heavy nuclides).

The weight of nuclear material needed to provide that quantity of fissile nuclides depends of the allowed BU. If the BU is for instance 45 GWjt-1, 21 tons are necessary. Then the reactor has produced 7.44 TWhe.

Typical figures for a PWR 900 MWe

Fuel sub-assemblies : UO2 (enriched up to 3.7 %), sintered fuel pellet, pins (Zr), pieces of structure (stainless steel, inconel) to accommodate 264 pins, command rods and monitoring. Same design if MOX is used (Udep up to 9% in Pu). T in pins ranges from 1600 to 500 °C over 0.5 cm.

Moderator and coolant: water under 155 bar/H2 (280 to 320 °C) containing B, pH around 7.

Core vessel : 157 sub-assemblies,13 m height, 4 m in diameter, thickness 20 cm, weight 320 tons. The active core is a cylinder of only 3,6 m height and 3 m in diameter.

Average = 1014 n cm2 s-1 or = 2.2 1019 n cm-3 (taking neutron with a speed of 2200 m s-1). In the same volume there are around 10 time more fissile nuclides (13.5 10 19)

Considerable radioactivity: 106 Ci, near EBq (1 EBq = 1018 Bq).

PWR Schematic view

Primary circuit

Secondary circuit

Water heater

Pump

Cooling water

Condensor

Generator

Control rods

Pressurizer

Vapor generator

Vapor Water

Primary pump

Vessel core

core

Containment building

PWR Sub-assembly

Fuelling (UOX based and renew of 1/4 core/year)

The first charge represents 72.5 tons of Uenr at 2.43 %, which has needed 316 tons of Unat.

Each year 40 sub-assemblies burnt at 41.2 GWjt-1 are replaced by new sub-assemblies of fresh fuel, enriched at 3.7 %. They represent 18.5 ton of Uenr which has needed 153 tons of Unat and 87 103 SWU (tail at 0.3%).

Typical figures in UOX SF are 95 % U, 1% Pu and 0.1 % for other actinides but they depend on BU

The yearly unloaded SF contains 800 kg of FP. For actinides (except U) the figures (calculated) are the following: 208.5 kg of Pu, 11.3 kg of Np, 12.5 kg of Am and 1.55 kg of Cm.

Quantities of FP and Actinides

The most important FP are stable nuclides (Xe, Zr, Mo, Nd, Cs, Ru, 70 % after 10 years) but mixed with very active radionuclides (Cs134 and 137, Sr90, Ce144, Ru106,…) and long-lived radionuclides (Zr93, Se79, Tc99, Pd107, I129, Cs135). The decay of the radioactivity of the FP is under control of Cs137 and Sr90 (T = 30 y). The quantity of FP is proportional to the BU, around 1% by GWjt-1

The quantity of actinides is not proportional to the BU. The isotopic composition of the initial Uenr is changed with the build up of U236 (0.5 %) and a still appreciable amount of U235 (1%) which make U in SF as energetic as Unat.

All actinides nuclides in appreciable amounts are very long-lived, except Pu238 and 241, Am242 and Cm244. They are also (except Pu241, - emitter) responsible of activity of the SF which is due for 55 % to Pu isotopes.

Energy comes from the fission of Pu for 30 %

Quantités de PF et actinides (sauf 238U) présents dans 1.13 t de combustible usé UOX1 (enrichi à 3.5% en 235U brûlé à 33 GWjt-1) 4 ans après sortie du réacteur

00,5

11,5

22,5

33,5

44,5

H Se Kr Rb Sr Y Zr Mo Tc Ru Rh Pd Ag Te I Xe Cs Cs Cs Ba La Ce Pr Nd PmSm Eu Eu Gd

Série2

Série1

UOX1

0

2

4

6

8

10

12

234235236238239240241242237241242m

243243244245

U Pu Np Am Cm

UOX1

134 135 137Produits de fission

H Se Kr Rb Sr Y Zr Mo Tc Ru Rh Pd Ag Te I Xe Cs Cs Cs Ba La Ce Pr Nd Pm Sm Eu Eu Gd

Np

kg

U

Pu

AmCm

kg

Radioactif (3.93 kg) Stable (34.226 kg)

MOX

0

5

10

15

20

25

234235236238239240241242237241242m

243243244245

U Pu Np Am Cm

MOX

Actinides Actinides

U Pu Am Cm U Pu Am Cm

kg

Safety of a reactor

It lies on “3 pillars»: security systems, barriers and a “Safety culture” of the operators. The security systems are independent and superfluous. They insure the integrity of the barriers. The barriers prevent dissemination of radioactive matter.

The 3 barriers are the cladding of the pins, the primary circuit which includes core vessel, stream generators and all the devices under pressure, and the confinement surrounding which enclose all the parts of reactor and utilities containing radioactive matter.

The operators have strict orders to run the reactor.

Shut down is assured by security rods (high a). The remaining power is few % of the initial value, due to the radioactivity of the core (a unique case in combustion)

Cooling is mandatory in every situation.

Two major accidents have occurred, Three Mile Island in 1979 (PWR reactor) and Chernobyl in 1986 (RMBK reactor).

In the first case some misinterpretations of indication have led to the partial emptying of the core vessel and a partial fusion of the core (Zr cladding heated to 1500 °C reacted with water to give H2). The surrounding confinement has prevented release of radioactivity to environment (around 1TBq of gaseous FP escaped).

In the second case the conception of the boiling water type reactor RMBK was special (control rods always in core to balance positive k/k void due to C as moderator and H2O as coolant, no surrounding confinement). Due to shortcoming of security systems during control tests, vaporisation of water increased to the point where the power reactor increased by 100 in few seconds. Nothing was possible to stop fission except that fuel exploded.

Gaseous water was reduced to H2 by Zr cladding, which gave a chemical explosion and the destruction of the reactor. All gaseous FP where immediately released to environment and the graphite burnt during 10 days because as all reactors with C as moderator the size of RBMK is large. The amount of radioactivity released is estimated to some EBq.

Today safety cases analysis of reactors are done with the objective to have a probability of 10-5 for an accident affecting the integrity of the core and 10-6 for an important release of radioactivity

Safety is a world-wide concern

World nuclear energy connected to the grid (end of 2004)

440 reactors were connected to electrical grids in 31 countries with a power of 366 GWe. They produced around 2620 TWhe, which represented 20 % of world electricity consumption but only 6% of total energy consumption on earth. Repartition of nuclear energy is variable. Some countries have decided to stop the use of nuclear energy and other to increase it.

Around 10 000 tons of SF were unloaded in 2004 requiring around 70 000 tons of Unat for refuelling. 28 reactors were in construction and others are planned according to forecasts on the economy growth of countries

Today PWR and BWR dominate the market (86%). They belong to the “Generation II reactors” lying on thermal neutron reactors fuelled with low enriched U (or Unat) and partially with Pu. “Generation III reactors” will have additional systems to improve safety

Table 7. World-wide nuclear reactors according to systems (end 2004)

Systems Reactors connected to grids MWe Number of reactors

PWR 204 441 214VVER 35 776 53BWR 82 550 93CANDU 19 972 39Magnox, AGR1 10 664 22RBMK 2 11 404 16FNR 3 1 039 3

Total 365 846 440

1 Moderated with C, cooled with CO2, U metal (Magnox) UO2enr (AGR) in Great Britain2 Moderated with C, cooled with boiling water, UO2enr (ex URSS)3 Fast neutrons, cooled with sodium, UO2dep, PuO2

EPR with a power of 1.6 GWe is designed for 60 years with a loading factor of 92 %. It will be fuelled with UOX or advanced MOX and the BU is expected to reach 70 GWjt-1. Its safety will lie on improvement on control command, resistance to seismic event and commercial aircraft crash and on recuperation of “corium” in case of core melting.

Between 1950 and 2004, 108 reactors (35 GWe) have been decommissioned and are for the major part dismantled or to be dismantled

EPRGeneration III reactor

Double containement building

Corium recuperation

Indor water tank

Redundant engineered safety system

Fuel cycle

The fuel cycle associated to a given reactor type consists of all the steps from mining of U to management of ultimate nuclear wastes. For all reactors the steps from mine to the preparation of sub-assemblies are the same. When SF are unloaded there are two strategies.

To consider the SF as waste (open cycle) or to consider the SF as a resource of fissile and fertile nuclides (closed cycle). It is a choice of countries to manage SF according to a given strategy, which lies on opposite arguments. The choice of the closed cycle leads to reprocess the SF. Large reprocessing plants are necessary (up to 1 000 t SF/year).

Reprocessing

About 1/3 of unloaded UOX SF have been reprocessed (75 000 over 250 000 t). U and Pu are separated from all other elements present in SF with a high yield (99. 8 %) and high decontamination factors (Purex process). Urep is stored (U3O8)

and Pu is converted in PuO2 and used to prepare MOX fuel. MOX SF is not reprocessed (storage of Pu)

All the elements from the SF (FP, Np, Am, Cm, 0.1 to 0.2 % of Uret and Pu) and chemicals added in process are confined in nuclear glasses and packaged in stainless steel containers (HL-LLW). The heads and ends of sub-assemblies and hulls are presently packaged in France as compressed metal in inox containers (ML-LLW). There are other wastes

The volume of reprocessing wastes is 0.5 m3 t-1 in France which is low compared to the 2 m3 of 1 ton of SF, but Urep must be stored and Pu recycled

NPP

Reprocessing plantUrep, Pu, wastes

Uenr, fuel fabrication

Unat (extraction, conversion)

PWR Fuel cycle

Urep Pu

UOXMOX

Upper cycle Reactors Back end cycle