Nuclear energy
Environment, 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 AZX ( (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
H1
Li6
B10
N14
O17
Cl35
Mo95, 97
Ag107, 109
In115
Xe131, 135 (T = 9h)
Cs133
Pr141
Nd142, 143, 145
Sm149,150,152
Eu151,153
Gd155, 157
Cr50, 53
Mn55
Fe, all isotopes
Co59
Ni58, 62
Reaction (n,)
0.330 (gives D)
Reaction (n,p)
Reaction (n, )
940 (gives T)
3840 (gives Li7)
1.80 (gives C14)
0.240 (gives C14)
43,6
14.5, 2.5
37, 90
115
90, 2.85 106
28
11
19, 330, 47
40.1104, 100, 200
9200, 370
6 103, 2.54105
16, 18
13
2.5
30
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.
137
97
%
U235
10
Tc99
1
Xe135
Zr95 Ru106
Sr90
I129
Cs137
Ce144
0
118
0.1
U235
Pu239
Fast neutrons
Kr85
0.01
+ 0.1 % T
0.001
Thermal neutrons
75
100
A
125
Induced fission with thermal neutrons
150
Log10 
4
resonances
3
2
Fast neutrons
Slow neutrons
1
0
-3
-2
U235
-1
0
1
2
3
Pu239
4
Log10 En(eV)
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)
t (s)
N/fission  (pcm) En(MeV)
10-3
Br87
I137,Br88
I138,Br89
I129, Br90
I140
Total
55.7
22.72
6.2
2.3
0.6
80.4
32.8
8.9
3.3
0.88
5.2
3.46
3.10
6.24
1.82
18
Th232
U233
U235
U238
Pu239
Am241
Cm242
21
142
123
257
75
650
4000
300
650
800
200
30
40
0.250
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,
thermal n
f
c
U235
38.8
8.7
U238
0.103
0.86
Pu238 2.4
27.7
Pu239 102
58.7
Pu240 0.53
210.2
Pu241 102.2
40.9
Pu242 0.44
28.8
Np237 0.52
33
Am241 1.1
110
Am242 159
301
Am242m 595
137
Am243 0.44
49
Cm242 1.14
4.5
Cm243 88
14
Cm244 1.0
16
Cm245 116
17
PWR MOX
epithermal n
f
c
12.6
4.2
0.124
0.8
1.9
8
21.7
12
0.7
24.6
28.5
9
0.5
12.3
0.6
18
0.8
35.6
126.6
0.5
0.96
43.1
1
33.9
27.5
31.7
3.45
7.32
13.1
5.4
FNR UOX MOX,
fast n
f
c
1.98
0.57
0.04
0.30
1.1
0.58
1.86
0.56
0.36
0.57
2.49
0.47
0.24
0.44
0.32
1.7
0.27
2,0
3.2
0.6
3.3
0.6
0.21
1.8
0.58
1.0
7.2
1.0
0.42
0.6
5.1
0.9
Z,N
p,i
p,i
p,i
i,p
i,i
i,i
p,i
p,i
Building of actinides
234 235 236 237 238 239 240 241 242 243 244 245 246
Cm
Main road
Am
85%
m
75%
Pu
71%
Np
U
n,2n
n,
75%
EC

n,f
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
s
1
2
3
4
5
6
7
H
1
Li
3
Na
11
K
19
Rb
37
Cs
55
Fr
87
8
9
10
11
12
13
14
15
d
Be
4
Mg
12
Ca
20
Sr
38
Ba
56
Ra
88
16
17
18
O
8
S
16
Se
34
Te
52
Po
84
F
9
Cl
17
Br
35
I
53
At
85
He
2
Ne
10
Ar
18
Kr
36
Xe
54
Rn
86
No
102
Yb
70
Lr
103
Lu
71
p
Sc
21
Y
39
La*
57
Ac**
89
Ti
22
Zr
40
Hf
72
Rf
104
V
23
Nb
41
Ta
73
Db
105
Cr
24
Mo
42
W
74
Sg
106
Mn
25
Tc
43
Re
75
Bh
107
Fe
26
Ru
44
Os
76
Hs
108
Co
27
Rh
45
Ir
77
Mt
109
Ni
28
Pd
46
Pt
78
Cu
29
Ag
47
Au
79
Zn
30
Cd
48
Hg
80
B
5
Al
13
Ga
31
In
49
Tl
81
C
6
Si
14
Ge
32
Sn
50
Pb
82
N
7
P
15
As
33
Sb
51
Bi
83
110
111
112
113
114
115
**
5f
*
4f
Th
90
Ce
58
Pa
91
Pr
59
U
92
Nd
60
Np
93
Pm
61
Pu
94
Sm
62
Am
95
Eu
63
Cm
96
Gd
64
Bk
97
Tb
65
Cf
98
Dy
66
Es
98
Ho
67
Fm
100
Er
68
Md
101
Tm
69
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
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
energy
Uenr (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 ai)
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, t, 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/t (P = exp 10 in
1 s, t = 0.1ms, keff = 1.001). Impossible to control the propagation
Fortunately there are  % of delayed neutron (coming as said from
some FP which are emitted t 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 t 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
Vapor generator
Secondary circuit
Pressurizer
Vapor Water
Control rods
Primary
pump
Generator
core
Condensor
Pump
Vessel core
Cooling water
Water heater
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 longlived 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.
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
kg
Produits de fission
134 135 137
 Radioactif (3.93 kg)
 Stable (34.226 kg)
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
Actinides
kg
Actinides
kg
Pu
Am
U
U
Pu
Am
Cm
Cm
Np
U
Pu
Am
Cm
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
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
VVER
BWR
CANDU
Magnox, AGR1
RBMK 2
FNR 3
204 441
35 776
82 550
19 972
10 664
11 404
1 039
Total
365 846
214
53
93
39
22
16
3
440
1 Moderated with C, cooled with CO2, U metal (Magnox) UO2enr (AGR)
in Great Britain
2 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
EPR
Generation 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 (HLLLW). 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
PWR Fuel cycle
Unat (extraction, conversion)
Uenr, fuel fabrication
Urep
Reprocessing plant
Urep, Pu, wastes
Pu
UOX
MOX
NPP
Upper cycle
Reactors
Back end cycle