• Abstract and Figures
• Public Full-text

1

Euro-Mediterranean University in Serbia Vrnjacka Banja, October 17-24, 2006

Nuclear energy Environment, sustainable development, renewable energy

Robert Guillaumont French Academy of Sciences

Nuclear energy is used to produce electricity in electric-nuclear reactors. The main problem to make nuclear energy safe is to avoid uncontrolled dissemination of radioactive matter. To discuss the topics of nuclear energy and environment, nuclear energy and sustainable development and 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 waste. The information is, as much as possible, self- supporting. The figures are generic or specific.

1 - Figures on the electric-nuclear energy

1.1 - Radioactive matter and nuclear energy

Basic definitions

A The probability that a nuclide ZX disappears at t, during dt, is dPt, t+dt(1) = ( + ) dt =  dt, where :  (t-1) is the radioactive constant of a spontaneous decay event,  (10-24 cm2 or barn, A b) is the cross section of a nuclear process in the incoming flux  of given particles on ZX ( (particles cm-2t-1). Theses parameters are independent of the oxidation state of X. In the following only some spontaneous events (, -, sf decays) and nuclear reactions by neutrons (n, ), (n,2n), (n,f) and , (,n) have to be considered. When a nuclide, i, disappears a new nuclide, j, (for , -, (n, ), (n,2n), ..), or several new nuclides j (for sf, (n,f), ..) appear.

For a given number N of nuclides the probability that Ni nuclides disappear at t, during dt, is given by the binomial distribution

Ni N-Ni Pt,t+dt(Ni) = [N!/Ni!(N-Ni)!] (dt) (1-dt)

for which average value is = Ndt . Except in few cases, N is always over and over 100 units and it can be considered as a continuous variable. Hence, among a population of N nuclides present at t, and during dt, the number of nuclides which disappear is - dN = =

Ndt, assuming that dN << N. It follows the well known law N = N0 exp-t. If N < 100, the binomial distribution apply.

For spontaneous decay ( = ) the half-life T is defined by N = N0/2 and it follows T = Ln2 = 0.7. The validity of the decay law dN << N leads to Dt << T, dt. The mean lifetime , is the t t t average time of the life of the nuclides,  = 0 tNdt / 0 tNdt = 0 exp-t dt = 1/ = T/0.7. As

 = 1,  is also defined by N = N0 exp –1 = 0.368 N0. Therefore for radionuclides one can speak of half-life or of mean lifetime to characterise long-lived or short-lived radionuclides. The activity of N radionuclides is A = N = 0.7/T (Bq = 1 decay s-1). The old unity of activity, 2 still used for very active matter, is the Curie (1 Ci = 3.7 1010 Bq = 37 Gbq, 1 Bq = 27 picoCurie)

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/. Otherwise the situation is complicated. The build-up kinetics of nuclides disappearing and appearing by decays and nuclear reactions are given, for each nuclide, as solutions of sets of differential equations, dNi(t)/dt = ji CijNj(t) – Cii Ni(t), giving the number of Nj(t) nuclides (nuclides cm-3) produced by all possible nuclear processes on nuclides i, the nuclides i disappearing itself by all possible nuclear processes and Cij or Cii being the yields depending of ,  and . There are computational codes to solve the systems of equations, when all  and  are known, like f or c respectively for neutron induced fission and for the capture of a neutron. Today coherent and updated international. “Charts” and “Data bank” exist. Tables 1, 2, 4 and 5 give  and T values of some pertinent nuclides.

Fission and other nuclear processes, microscopic aspect

Nuclear energy comes from complicated and competing nuclear processes in a given high radioactive matter submitted to a high neutron flux and containing fissile (f > c) and fertile

(c > f) nuclides, these relationships being verified over all neutron energy (En, ranging from kT to few MeV). Slow neutrons have En < 1 eV and fast neutrons have En > 0.1 MeV. Matter, which does not contain fissile/fertile nuclides (inert matter), generally surrounds the matter where nuclear energy is released and makes possible the control of the release of nuclear energy ‘see later). Inert matter contains nuclides with appropriate c.

The main source of nuclear energy is the fission (possible when Z2/A > 37) of even and odd A isotopes of U and Pu induced by slow (A odd, Z even) and fast (A even, Z even) neutrons (mainly if En > 1 MeV). Neutron is absorbed by the nucleus (A,Z) and gives an excited nucleus (A+1,Z). If En is low the excitation energy given to the Nodd +1 = Neven nucleus by parity effect (pairing of neutrons) leads to large vibrations, which end by the fission of the nucleus/nuclide, it is to say to a redistribution of nucleons and electrons in new nuclides. When N is even the excitation energy is not enough to induce the fission of the Neven + 1 = Nodd nuclide. When En is high (fast neutrons) the probability of fission of the excited nucleus is not sensible to parity effect. Fission yields depend on the cross section function f = f(En).

Some odd-odd nuclides are fissile, like Am242. The cross section f depends of the speed, v, 2 of the neutrons. It decreases as 1/v for thermal neutrons (kT = 1/2 mnv =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). Neutrinos are also emitted. 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 - decay, giving daughters. FF and these daughters are called fission products (FP). The A and Z values of FP range over large limits, 30

The release of energy of a fission reaction is around 200 MeV. The FF take 166 MeV as kinetic energy. The energy of the  rays is 8 MeV and the prompt neutrons have 5 MeV of kinetic energy (En = 2 MeV, <> = 2.5). The neutrinos take away 10 MeV. Kinetic energy of FF is lost through inelastic coulombic interactions with the electrons of the matter.  rays loss a part of their energy through compton and photoelectric effects, but can escape the fissile/fertile matter. The bêta decay of FP (and associated  emission) is 29 MeV (10 for electrons, 9 for  rays and 10 for neutrinos). The contribution of these processes to energy given by fission is around 15 MeV because of the loss of neutrinos energy, but it is not instantaneous. All these figures are average values.

The fast neutrons, which do not escape the fissile/fertile matter, are trapped by some heavy nuclides, mainly U238 and U235 (and later by heavier nuclides) and give some fission. They do not loss much kinetic energy by elastic collision because the A values of the heavy nuclides is too high. If slow neutrons are present (as the result of their slowing down in the surrounding inert matter) nuclear reactions (n,) occur, like U235(n,)U236 or nuclear reactions (n,) followed by bêta decays, which produce actinides, for instance U238(n,,2-)Pu239. Actinides have Z > 98 and are  emitters. Very heavy actinides can decay by sf, like Cm244. There are also some (n,2n) reactions giving isotopes of the target nuclide, like Pu239(n,2n)Pu238 or some (n,2n,-) process, like U238(n,2n,-)Np237(n,2n,-)Pu236. There are in addition (,n) reactions on light nuclides, if present. All the processes not directly linked to fission contribute only for few MeV to the energy dissipated in the radioactive matter. The total energy associated to one fission is, in average, around 200 MeV, counting the neutrons energies En. Figure 2 gives the main nuclear processes of build up if actinides.

Fission energy, macroscopic aspect

Depending on A value, (235 to 241), for the fissile nuclides and according to the relationships: 1 eV.molecule = 23.0609 kcal.mole = 96.5098 kJ.mole and 1J = 0.2778 kWh, 1g of each of these nuclide give 8.00 1010 to 8.21 1010 J. For a given fissile/fertile matter the exact value depend on the isotopic composition of the fissile nuclides and on the cross sections of fission. Taking 8.2 1010 J (A = 235) as a scoping value shows that 1g of fissile nuclide give 34.3 107 kcal or 22780 kWh (or 1.8 tep), which is over the energy given by the combustion of 1 g of any material. One speaks about the « Burn Up » (BU) of a nuclear material as the energy (MWj) given by unit of weight (t). 1 MWjt-1 correspond to the release of 8.64 1010 J by ton of IHM (initial heavy metal), it is to say 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). One can accept the equivalency 1 MWjt-1 = 1.053 g of fissile nuclide = 1.053 g of FP. Here and in the following the weights are given in IHM according to the use in nuclear energy considerations. In a material burnt to 45 GWjt-1 about 47 kg of fissile nuclides have been transformed into FP.

The fission energy in power electric-nuclear plant is made available through transformation in heat, heat in work (reactor) and work in electricity (electric facility). Assuming a total yield of 33 % to have 1 MW electric power (MWe), 3 MW thermal power (3 MWth) are required, which needs each second the fission of 0.0365 mg of fissile nuclides. The electric power of a 4 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. 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. Loading factor and BU depend of the reactor and on the electric demand on the grid. It is convenient for rough estimations to adopt some equivalencies. For instance in France in 2004 one TWhe has needed 2.6 tons of nuclear fuel. When BU of SF was 33 GWjt-1 the equivalency was 1 TWhe/year = 3.5 tons/year (each year 21 tons was unloaded for 6 TWhe). All figures in nuclear energy can be evaluated with respect to TWhe or to tons of fuel.

1.2 - Nuclear reactor and spent fuel

Release of fission energy

As an induced neutron fission produce  neutrons a continuous process of fission of the fissile nuclides can be established providing the strict control of the multiplication of neutrons to a one to one ratio. If it is not the case the fission reactions lead to a so sudden high local release of energy that an explosion occurs or the fission reactions cannot propagate. Most of the nuclear reactors are based on fissions induced by thermal neutron whose energy is about kT. 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. Few reactor do not need moderator, there are called fast neutrons reactors (FNR).

The natural radioelements (elements for which all isotopes are radioactive or are radionuclides) 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 fissile nuclide which has

f > c whatever the neutron energy is (fissile nuclide). U238 and Th232 are fertile nuclides because the processes (n,) followed by - decay give Pu239 and U233, which are fissile nuclides. Consequently natural U (Unat) is the main primary source for the nuclear fuel. Thorium can be only used as fertile element when associated with fissile isotopes of U or Pu. So nuclear energy lies, up to now, practically on Unat. 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.

The moderators are materials which contain nuclides of low Z and low absorption cross section (a) to optimise both the energy losses by elastic collision with nucleus and the thermal range of neutrons: H2O, D2O, C (as graphite). Around 20 collisions to bring fast neutrons to kT are necessary with H, 25 with D and 120 with C but more than 2000 with U238. The respective ranges are 2.8 cm (H2O), 10 cm (D2O) and 5.2 cm (C). Thermal neutrons give a kind of “neutron gas”. In this case (cm2s-1) can be expressed as (cm2v) 1/2 with v = (2kT/mn) . Over H, D, C, O, H is the better moderator but has a quite high a value (0.39 for (n) reaction). In an elastic collision n-H half of the energy of the neutron is lost.

Coolants are materials which contain nuclides with the same characteristics as moderators: H2O, D2O, CO2, He. Unat cannot be used with H2O as moderator and coolant because H2O absorbs too much the neutron. Heavy water must be used. As D2O absorbs less the neutrons and because elastic collision of a neutron with D is less effective than with H the neutrons 5 have a longer range and such based reactors have large dimension (CANDU reactors). Another way is to use C as moderator and pressurised CO2 as coolant (UNGG reactors). With U enriched in U235, 3 to 5 %, (Uenr) the right conditions to run a reactor can be found with H2O. The reactors PWR or BWR are the most common. Of course they need to have plants to enrich Unat, which consume large amount of electricity. To get 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. It is counted in SWU (separation work unit). 1 SWU = 2400 kWh. So enrichment needs 3 103 to 5 103 SWUt-1. Udep, is stored. In FNR the coolant is Na (transparent to neutron). Pu has f values for fast neutrons lower than for thermal neutron, f < c and as c

Pu239 > c U235 they need large amount of fissile Pu nuclides.

Control of fission energy release

The fission reaction can be propagated if one neutron issued from one fission can give one additional fission, the (-1) other could being lost. It means that in an idealised infinite medium allowed for the propagation of the fission reactions the multiplication factor k = i i i ni f / ni a must remains always equal to one. 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 summation is extended to all the different nuclides of the medium. k depends on identified factors of which some can be controlled by using nuclides with strong c (control nuclides) or by adjusting the moderation ratio (nuclides of moderator over fissile nuclides). In a real system there are losses of fast and slow neutrons and an effective multiplication factor, keff = k.F, must be considered. If keff = 1 the reactor is critical and it is sub-critical if keff < 1. The -5 reactivity is  = keff – 1/keff. At any time the difference to  = 0 (keff = 1) is measured in 10 unit, called pcm (percent per mil). The time, , between two neutrons generations, g and g+1, is very short, some milliseconds for thermal neutrons and some microseconds for fast neutrons. The increase with time of the power of the reactor is such that P = P0 exp (keff-1)t/. This comes from the differential increase of neutrons, dn/dt, between two generations of neutrons: n(t+) – n(t), which leads to dn/dt = [keffn(t) – n(t)]/ and to n = n0 exp (keff-1)t/. It can be easily seen that any action to control P (or keff) is impossible. For instance with  =

0.1ms, keff = 1.001 the power would be exp 10 time P0 in one second !. Fortunately there are  % of delayed neutron coming from some FP, which are emitted  seconds after the prompt neutrons of fission. The safety of a reactor can be, then, insured by controlling the number of delayed neutrons with nuclides absorbing neutrons (high c). 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. This parameter depends on the fissile nuclides (Table 3). For U235, d is equal to 635 pcm. It decreases when A and Z increase and the control of the reactor is more difficult. As  % Pu239 <  % U235 operation of FNR is less easy than PWR.

There are some phenomena which contribute, or not, to the safety. If T increases the densities of liquids (moderator/coolant) decrease as well as the slowing down of neutrons, which are less effective for fission, and the moderation ratio also decreases. If T increases in fuel, neutron spectra change and f or c increase/decrease, particularly for the range of energy of resonances (of about 0.1 eV large for En > 10 eV). This effect is called Doppler effect because an increase of T increases the speed of U238 nuclides, which appear to neutron with enlarged 6

-1 resonances. For U238 c increases (keff/keff = -2.5 pcm°C ) but for Pu239 it is f. To keep the benefit effect of U238 only some % of Pu can be mixed, or accepted, with U.

The control of reactivity is done by different type of control rods made of nuclides absorbing thermal neutrons. Some have a slow movement. When starting the reactor, they are in the nuclear core and, as the fuel reactivity decreases (around 1000 pcm.month-1), they are slowly removed to compensate this natural evolution. Some with a rapid movement compensate other effects. Shut down of a reactor is realised by rods, which can be very quickly introduced in the core, or by injection of B in liquid moderator. H3BO3 added in the moderator when starting a reactor is a consumable poison The boron helps to control the excess initial reactor reactivity which is at its maximum.

As long as fission and nuclear processes occur the FP and actinides accumulate in the nuclear fuel and the conditions for keeping keff = 1 are finally not longer possible. FP as some lanthanide isotopes and particularly Xe135 are poisonous (Table1). The decrease of reactivity becomes difficult to compensate. Other factors occur as physico-chemical transformations, which change the properties of the fresh fuel and those of the alloys of the devices where it is encapsulated (cladding). The fuel has to be renewed. The unloaded fuel is called spent fuel (SF). Its average composition depends of the type of reactor and of its BU. As it is highly radioactive, due to short-lived radionuclides, and warm, it goes to a pool for deactivation during several years.

Figures for a typical PWR

Most of the electric-nuclear reactors in the world connected to electrical grids are PWR or BWR. Schematic view of a PWR is given in Figure 3, which shows the main components. Typical figures for the most common French “ 900 MWe PWR” are the following (electric power varies from 880 to 920 MWe).

The fuel UOX is enriched UO2 (2.43 to 3.7 % in U235 depending of the initial or recharge load, 3,5 % in average). The fuel MOX, when used for reload, is a mixture of depleted UO2 and of PuO2 (7.5 % in Pu in average but ranging from 3 to 9 % depending where MOX is located in the core). The fresh fuel is sintered in pellets (10 mm height, 8.5-9 mm diameter), which are put in pins under a He atmosphere. Pins are hermetic zircaloy tube, 4 m long and 10 mm in diameter, which contain each 2.5 kg of fuel for 0.5 kg of zircaloy (1.5 % Sn, 0.3 %

Cr,Ni, Fe). Zircaloy is neutron transparent (c = 0.18 b). A total of 264 pins (17x17) are put together in a sub-assembly (5 m long, 21.5 x 21.5 cm)which amount to 140 kg of Zr, 10 kg of stainless steel, 6 kg of inconel (Figure 4). Each contains 524 kg of fuel or 461 kg of U. A sub- assembly is the basic item for all handling. A sub-assembly is designed to receive control rods and monitoring devices. A total of 157 sub-assemblies are necessary to run the reactor. They are put in the core vessel designed to control all the functions necessary to operate the reactor. The core vessel is 13 m height, 4 m in diameter, has a thickness of 20 cm and weights 320 tons. The active core of the reactor is a cylinder of only 3,6 m height and 3 m in diameter. The gradient temperature in the pellet is very high (around 100 °C mm-1 from 1600 °C to 500 °C). The water of the primary circuit is under pressure to avoid ebullition (in 280 °C, out 320 °C,155 bar under H2) and flow through the sub-assemblies and pins pushed by pumps. The water contains H3BO3 (B10, 20%, c = 3800 b, B11, 80%, c = 0) and its pH is continuously adjusted to pH 7 (at 300 °C) with LiOH (Li6, 7.5%, c = 940 b, Li7, 92%, c = 0). The value of pH is the best to optimise the corrosion of metallic components and to hydrolyse the corrosion products (Fe, Co, Cr, Ni, Zr). Radiolysis of water is high ( rays and neutrons loss 7

2% of their energy). O2, H2, OH and H are produced despite the fact that H2 reduce the water radiolysis. O2 and H2 are recombined to water. Typical value for the average neutron flux is 1014 n cm2 s-1 or  = 2.2 1019 n cm-3 taking neutron with a speed of 2200 ms-1. In the same volume there are around 10 time more fissile nuclides (13.5 10 19).

The artificial radioactivity which is build up in the water comes from induced neutron nuclear reaction on O isotopes and on the elements dissolved (C as CO2 N as N2, B and Li) like B10(n,2)T and Li6(n,)T, from gaseous FP which diffuse through the pins (Xe133 and 135, Kr85, I 131 and 129, Cs134 and 137, T issued from ternary fission) and from activated corrosion products (Cr51, Mn54, Fe59, Co58 and 60, Zn65). The water can reach an activity of 1 to 10 Ci m-3 in T as HTO. It is treated on ions exchangers, the gases are stored for decay and then released as effluents.

Neutrons induce nuclear reactions mainly in all the metallic materials of the reactor (Co60 in heads and ends of sub-assemblies for instance).The artificial activity, which is build up in the fuel, is enormous. The total activity of a reactor amount tens of 106 Ci near EBq (1 EBq = 1018 Bq).

A French 900 MWe reactor is designed for 30 years but its exploitation will be extended to 40 years or more. Its yield ranges from 31.7 to 33 %. The first charge represents 72.5 tons of Uenr at 2.43 %, which has needed 316 tons of Unat. Then fuel management depends on BU. Typically each year 1/4 of the core is unloaded (40 sub-assemblies burnt at 41.2 GWjt-1) which have lasted in reactor 48 months. These sub-assemblies 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%). When MOX is used to reload a reactor in the proportion of 30% the figures are different. Fissile isotopes of Pu decrease the need of U235.

As f of Pu239 and 241 are less than of U235 the % of Pu in MOX fuel is about 7.5 % in average. The yearly recharge of 1/4 of the UOX core and 1/3 of the MOX core is of 28 sub- assemblies of UOX and 16 of MOX needing 7.4 tons of mixed oxide (0,55 tons of Pu and 6.85 of Uapp). The MOX lasts 38 months in the reactor. The economy on Uenr is 5.6 tons. The BU of the unloaded MOX is 33.8 GWjt-1. All these figures show that the yearly unloaded SF for a typical UOX-PWR contains 800 kg of FP and 720 kg for a UOX-MOX-PWR. For actinides (except U) the figures are the following: 208.5 kg of Pu, 11.3 kg of Np, 12.5 kg of Am and 1.55 kg of Cm (UOX-PWR) and 125.5 kg of Pu, 5.36 kg of Np, 3.65 kg of Am and 0.31 Kg of Cm (UOX-MOX-PWR).

The composition of the spent fuel depending of BU is given in Tables 4 and 5. After the decay of short-lived radionuclides the main features are the following. 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). All FP in SF are “artificial elements” with isotopic composition different from the natural ones. 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. It is not the case for actinides. Typical figures in UOX SF are 95 % U, 1% Pu and 0.1 % for other actinides. 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. Among the neutrons emitters by sf, 8 the most active is Cm244 (97 % of some 108 ns-1, including the additional neutrons coming from (,n) on oxygen which amount to 5 %).

The management of sub-assemblies in a reactor is subject to changes depending on a continuous improvement of fuel performance to increase BU.

All these figures (Table 5) allow to understand where fission energy comes from. For instance in 1 ton of SF burn at 45 GWjt-1, 47.39 kg of fissile nuclides have disappeared. According to the composition of SF, 30.314 kg of U235 (decrease of enrichment from 3.7 to 0.7%) and 33.90 kg of U238 (decrease by nuclear reactions and fission) have disappeared. As average values of c/f +c = 0.18 for U235 and c/f +c = 1.12 for U238 only 27.70 kg of U235 and 3.6 kg of U238 have been burned by fission. The difference between fissile nuclides and those of U is 16.16 kg. It represents the contribution of the Pu isotopes, which amount to 35 %.

1.3 - Safety of a reactor

The safety of a “safe reactor” lies on “3 pillars»: security systems, barriers and a “Safety culture” of the operators. The security systems are independent and superfluous, in accordance with the principles of nuclear safety. They insure the integrity of the barriers. The barriers prevent dissemination of radioactive matter according to the principle of “defence in depth”. The operators have strict orders to run the reactor. Safety is also imperative for economical reasons.

The power of a PWR reactor, for instance, is adapted to the demand in electricity by the quantity of stream directed to the alternators. As previously said the reactivity  in controlled by rods, which move slowly and automatically in the core, in addition to the passive use of nuclide (like B in water) to compensate natural decrease of . They contain nuclides with medium c values for thermal neutrons. In general if some disturbances occurs  decrease by itself, as well as the power of the reactor (intrinsic safety). Nevertheless in the case where something is wrong and necessitate the shut down of the reactor security rods, which have high “antireactivity” (nuclides with high c), fall instantaneously in the core, reducing the power to few % of its initial value. This remaining power is due to the radioactivity of the core. The 3 barriers are the cladding of the pins, the primary circuit that 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. Accidents can come from a bad cooling of the core or from a sudden increase of , and it is why there are systems to control the cooling even after the shut down.

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 (cladding heated to 1500 °C have 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. Moderator was graphite and H2O as cooling liquid circulated in tubes under pressure where it was heated by the fuel and partially vaporised. By conception k/k due to void following vaporisation of H2O (transition liquid-gas give void) was positive. Indeed H2O vapour absorbs less the neutron than liquid H2O while their slowing down to thermal energy due to C continues. It results an increase of  and to control the reactor the command rods have to be always in the core limiting the possibility of action. Finally the reactor was not enclosed in a surrounding 9 confinement. Due to shortcoming of security systems during control tests at a small power, vaporisation of water increased to the point where the power reactor increased by 100 in few seconds and nothing was possible to stop fission except that fuel exploded. Gaseous water was produced and reduced to H2, 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 was large. The amount of radioactivity released is estimated to some EBq. Today RMBK reactor are no longer in use as they were in 1986.

These accidents have led to international exchanges about reactor safety (WANO - World Association of Nuclear Operators, AIEA). 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. Nuclear safety is a world concern. There are common accepted rules whose application is controlled. 50 countries have signed the international convention on Nuclear safety (1996) and 30 on the management of SF and Nuclear wastes (2001)

1.4 - World nuclear energy connected to the grid

There are many possibilities to combine fissile nuclides, fertile nuclides, moderator (as well as the moderation ratio) and coolants. Using nuclear energy to get electricity needs to have the maximum of fission events and a long life of the fuel. Consequently fertile nuclides have to be used to renew the disappearance of the initial fissile nuclides. Only a limited number of types of reactors have reached industrial level. At the beginning of the use of the nuclear energy to produce electricity (1940 in USA and Canada, 1950 in France and Great Britain) the choices have been, according to resources at the time, Unat and C or D2O. Then enrichment of Unat has opened the possibility to use H2O.

At the end of 2004, 440 reactors were connected to electrical grids in 31 countries with a power of 366 GWe (Table 6 and 7). They produced around 2620 TWhe, which represented 20 % of world electricity consumption but only 6% of total energy consumption on earth. It is interesting to note that renewable energy contribute to 3.8 % (biomass and waste, 1.1 %, hydraulic, 2.3 %, geothermy, sun and wind, 0.5 %) and that fossil fuels contributes to 79.5 % (oil, 34.0 %, coal, 23.5 %, gas 21.1 %). Repartition of nuclear energy is variable (30 % for more than 15 countries, near 80 % in Lithuania and France but 2.2 % in China). Around 10 000 tons of SF are unloaded requiring around 70 000 tons of Unat (figures for 2005) for refuelling. Some reactors are in construction (28 in 2004) and others are planned according to forecasts on the economy growth of countries. The actual world-wide situation is well known. The tendency is as follows: in USA the renew of nuclear energy announced some years ago is still waiting, despite numerous initiatives, in European Union the tendency is uncertain and it is favourable in Asia. Some programmes are very ambitious (China). In the 15 countries of the European Union the capacity of nuclear energy is 122 GWe (137 reactors PWR or BRW, 35 % of electricity). Germany has decided in 2000 to stop progressively nuclear energy (19 reactors) on the basis of a 32 years for the life of reactors (end in 2022). Great Britain has rather old reactors and does not push for an immediate renew and in 2020 nuclear power will be reduced. Sweden has decided in 1980 to close all its reactors (12) before 2010 but this date is under reconsideration. Spain does not consider an extension of its fleet. Belgium has adopted in 2002 a law to leave the progressively nuclear energy on the basis of a 40 years for the life of reactors. Their closure should be effective between 2015 and 2025. The Netherlands must close its reactor. Italy has abandoned nuclear programme in 1980. Finland 10 and France have launched the construction of one EPR in each country. Extension of European Union to 25 countries will increase by 70 GWe its nuclear electric capacity (161 reactors). The new nuclear countries keep the nuclear option open. Switzerland has recently (2003) decided to do the same.

As already said reactors using Unat are large because of low content in fissile nuclide. Those cooled with CO2 or He gas under pressure and moderated with C must use metallic U to increase density of fissile nuclides with a high moderation ratio (C/U = 60). To avoid fusion of nuclear fuel BU is rather low (6 GWjt-1). It is why they have been supplanted by water moderated-cooled reactors, which are less large. When moderated-cooled with D2O, UO2 can be used as fuel and BU is higher (8 GWjt-1). They can be loaded and unloaded without stopping of the reactor. They have been developed by Canada since 1984 (CANDU). Unat fuelled reactors cannot reach large electric power (500 MWe max). Great Britain has developed in addition to Unat-C-CO2 reactors particular C-CO2 reactors (AGCR) fuelled with low enriched UO2 (2.5 %)

Today PWR and BWR dominate the market (86%). They are coming from US submarine reactors. Use of water as moderator-coolant leads to fuel it with Uenr with a moderation ratio around 2.5. (H2O/UO2). The technologies of PWR, BWR and CANDU have more than 30 years of experience. They belong to what is called the “Generation II reactors” lying on thermal neutron reactors fuelled with low enriched U (or Unat) and partially with Pu.

The projects of “Generation III reactors” (600 to 1500 MWe) considered by France- Germany, USA, Japan, Canada, Russia will have additional systems to improve safety, like the EPR (Figure 6) . EPR with a power of 1.6 GWe, to be economically competitive (price of kWhe around 3 cents of Euro with 8% of interest), 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. New Japanese reactors are close to generation III type reactors.

The technology of FNR Na cooled lies on operating 6 reactors during 20 years (250 to 1200 MWe). Only 4 are in use. Two high temperature reactors HTR (250, 300 MWe) cooled by He and using a special fuel have been operated but stopped (economy, some technical problems). Two experimental HTR reactors are in operation. Due to their interest (high thermodynamic yield and low power, possibility to burn Pu) projects are under development. Finally there are also projects included in the “International Forum Generation IV” (GIF), expected to be implemented in the second half of the century. Ten countries participate to GIF and 6 types of “Generation IV reactors” are considered, based on thermal or fast neutrons with open or closed fuel cycles. All will have a high temperature coolant (above 600 °C). They will earn resources in U, minimise wastes and led to less proliferation and less costs

Nuclear reactors can be seen as belonging to different systems with regard to Pu, those which produce Pu (generation II and III), those which burn Pu (HTR), those which give as much Pu as they burn or give more (FNR).

The different type of reactors have born in different countries for historical reasons but today there is a world-wide competition (technology, safety, wastes). 11

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

1.5 - Strategy for the back end fuel cycle

Running nuclear reactors needs nuclear fuel. The fuel cycle associated to a given reactor type consists of all the steps from mine of U to management of ultimate 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) (Figure 6). 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 1000 t SF/year). The countries, which have chosen a closed cycle, are France, Great Britain, Japan, and Russia. Some other have partially reprocessed their SF in France and Great Britain (Germany, Belgium, Switzerland, Netherlands)

1.6 - Reprocessing

About 1/3 of unloaded UOX SF have been reprocessed (75 000 over 250 000 t) but the tendency is going down. U and Pu are separated from all other elements present in SF with a high yield (99. 8 %) and high decontamination factors (104 to 105 for FP in Urep and Pu, 10 microgramme of Pu/kgU). The “Purex” process is used in all reprocessing plants. It consists to cut the pins in pieces and to dissolve irradiated UO2 in hot concentrated HNO3, to clarify the solution and to extract U(VI) and Pu(IV) in a solution of TBP in an aliphatic solvent and then to strip Urep (as U(VI)) and Pu (as Pu(III)) in separated acidic aqueous solutions. Urep is converted in U3O8 and stored and Pu is converted in PuO2 and used to prepare MOX fuel. The heads and ends of sub-assemblies and hulls are MLLLW. There are presently packaged in France as compressed metal in stainless steel containers. All the elements from the SF (FP, Np, Am, Cm, 0.1 to 0.2 % of Uret and Pu) and chemicals added in the Purex process are HLLLW. They are confined in nuclear glasses and packaged in inox containers. In addition there are secondary MLLLW coming from used technical materials and decontamination of primary aqueous and organic effluents. Some are managed with the previous ones. 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. Reprocessing of MOX SF with the Purex process is possible. Today it could be only reprocessed if it is mixed with UOX. Reprocessing of irradiated UOX at low BU (0.1 GWjt-1) is mandatory to get military Pu (Pu239). New methods of reprocessing and MOX fuel fabrication (including all actinides) are under investigation using both aqueous and dry chemistry. They will be a key issue for a future of nuclear energy (see later).

The world-wide capacity of reprocessing are around 5 700 tons of UOX SF.

1.7 - Advantages-disadvantages of safe nuclear energy

The advantages are the following.

Nuclear energy is a concentrated energy source (adapted to megapoles), with modest land use. Power station can be sited on the sea avoiding use of fresh water and coolant towers. Accordingly it is not very dependent of weather or climate. 12

A well proved technology, which can be still largely improved. More than 11 000 equivalent years of operating nuclear reactors have led to master nuclear energy. Due to continuous high level performances the life of reactor is increasing (from 30 to 60 years, depending the countries) as well as the annual load ratio (70 to 90%, 80 % in average), decreasing the cost of kWhe paid by customers (a factor 2 in 30 years). The BU of fuel is also increasing, earning SWU and fissile nuclides and decreasing unloaded quantities of SF by a factor of 1.5 to 2 (today a BU up to 60 GWjt-1 is reached). At the same time the quantity of Pu in SF decreases (35 to 25 kg/TWhe from 35 to 60 GWj t-1) but the build up of Pu238 and Cm244 increase (a factor 2 from 35 to 60 GWj t-1). This is not without consequences for a reuse of Pu and for waste management (change in isotopic composition of Pu, increase of the release of heat and neutrons emission by sf)

Important and diversified sources of natural and man-made fissile nuclides (Pu isotopes). Unat remains the main source of all the fissile nuclides. The low quantity of recycled Pu (or U rep) does not change anything (around 150 tons of MOX are used in the world versus 10 000 tons of UOX). U ores are widespread and abundant without competing uses and it is easy to store. The resources are in Australia (24%) Kazastan (17%) Canada (13%) and South Africa (9%) Today the resources at a price of 80 Euro/kg U are estimated to 2.3 Mt and to 1.6 Mt at 160 Euro/kg. World-wide consumption for 2500 TWhe is around 0.07 Mt/year. So more than 50 years of production at the present level of production can be achieved in the ranges of a prices indicated. In other situations the price will be increased (see later) although coast of U constitutes a moderate fraction of the price of kWhe. If U price was 10 times more the price of nuclear kWhe would increase of 40 %.

A low cost of kWhe. The total cost of kWhe produced by large nuclear reactors, including provisions for decommissioning and waste management has been calculated according the financial rules in use in “full life cycle cost” analysis. It is about some cents of Euro (3 for the last estimation).

A very low radiological impact for the environment and no release of CO2 (about 5 to 10 g of C/ kWhe or 40 g of CO2/kWhe against 1 kg for a fuelled coal electric plant of 600 MWe). Nevertheless as it stands nuclear energy does not contribute to reduce significantly CO2 emission (6% of the world energy consumption). It avoids around 2 Gt of CO2 emission per year.

A worldwide system of protection for workers. The rules are proposed by the ICRP and applied by country under the inspection of safety authorities. The rules lie on the principles of precaution and on the ALARA principle. There are many regulatory requirement.

Extension of the present technology to produce electricity (with reactors of 300 MWe) and high temperature heat, which could be used in metallurgical processes, and to produce hydrogen. New technology using all U isotopes (and Th) could provide fission energy over a very long time.

The disadvantages of nuclear energy are the following

A high investment costs with a long return of money (10 years for construction and a large amount of money in case of accident, converting financial assets in liabilities (TMI for instance). Immobilisation of capital has a cost. Financial market for investment increase rate interest to cover perceived risks. 13

A difficulty to follow variable electricity demand. High power reactors can only produce electricity on the grid at a «base load generation ».

An accumulation of civil Pu. In average the world-wide production is 30g kWh-1(75 tons a year). The status of Pu is peculiar according of its use as a fuel or its management as waste. Today Pu can be only recycled once in PWR and BWR (MOX fuel). As Pu239 has a lower  % of delayed neutron than U235 % the quantity to be introduced in reactor is limited to 12 % (30 % of MOX sub-assemblies in a 900 MWe PWR initially loaded with UOX). The isotopic composition Pu MOX SF does not allow a second recycle (too much even-even nuclides which decrease the quantity of fissile nuclides by thermal neutrons). Single recycle of Pu leads to a reduction of the stock of Pu up to 10 to 15 % (3 kgTWhe-1) with regard to open cycle. But the production of minor actinides increases and this has consequences on management of spent MOX fuel. Finally use of Pu necessitate to reprocess SF with accumulation of stockpile of a new kind of U, Urep (1% in U235 and 0.5 % in U236). The worldwide stockpile of Pu is increasing. Attached to the accumulation and separation of Pu from SF the question of dissemination of fissile material is central. Pu is highly radiotoxic. The quality of civilian Pu for making high yield nuclear engine is very poor, nevertheless separation of Pu from SF is considered as encouraging nuclear weapons proliferation.

A production of nuclear wastes difficult to manage whatever they are SF or packages of wastes from reprocessing. There is no technical problem in the short term. SF are stored in pools or in canisters naturally cooled. The LLW, which contain only short-lived radionuclides (PF and PA), are disposed of in ground or underground repositories. After 300 years their radioactivity will be very low. The HLLLW and MLLLW, which contain long-lived radionuclides, are put in interim storage. This storage licensed for some decades could be extended over centuries in special facilities. Nevertheless, « packages of wastes » (SF or reprocessing wastes) should be, finally, disposed of in geologic formations. This is the only way to confine radioactivity over period of time in accordance with the half-life of actinides and long lived FP. Modelling of the return of radionuclides to the biosphere from such disposal in a far future is, of course, stamped with uncertainties. Not all the experts are satisfied about the safety of disposal methods proposed over the long time to be considered. The final destinations of nuclear wastes are the matter of hard discussions. Management of these wastes is a key issue for nuclear energy.

A controversial appreciation by citizen against a rather new industry (about 30 years old) . There are opposite positions between anti-nuclear people and nuclear technocrats. The civilian nuclear is issued from the military nuclear and has kept during a long time its practises. Chernobyl is always in mind and has shown that a very low risk can have tremendous consequences. The spectrum of radioactive terrorism has recently appeared. Local oppositions on disposal of radwastes are sometimes intense. Radioactivity and dose and effects of low doses on long term are difficult to understand (see later). In short the hazard of nuclear energy comes from an exposure to ionising radiations, appreciated by the calculated “dose” The dose depend on scenarios of exposure. The risks depend on the dose. A reference point is the natural dose. Real efforts, but too recent, have been done to give an objective information on nuclear problems, but in general the dialogue is difficult to be established except in some countries (Sweden, Finland).

In conclusion the weakness of civilian nuclear comes from societal problems rather than from technological problems. They are linked with the back end of nuclear cycle. Waste 14 management is the main problem to be solved and many countries are doing research in this field.

2 - Nuclear energy, environment and society. Is this energy compatible with sustainable development?

Nuclear energy does not release chemical combustion products to atmosphere but has some impacts on environment mainly by the release of radioactive effluents and heat. The local radiological consequences on living species, including man, are usually low to very low, if compared to those due to some radioactive natural environments. It is the same situation for heat (reactor) and chemical pollutants (utilities of nuclear cycle) regarding other possible transfers from man-made systems to atmosphere, hydrosphere and geosphere. Of course in the case of accidents the situations have been or should be different. The links between nuclear energy and society are complex and will not be discussed in this paper otherwise than through the concept of radiological dose and associated risk (see later).

Sustainable development requires to meet several criteria, both technical and societal. In this section these criteria are essentially discussed in the context of the pursuit of the present technology of reactors (thermal neutrons fuelled with Unat or low enriched Unat and partially with Pu), according to a moderate increase of the use of nuclear energy. Indeed the time to move from Generation II (present) and III (coming soon) reactors to Generation IV (after 2050 ?) will be over few decades, even if research and development of prototypes is strongly supported. Up to 2050 nuclear energy will lie on U/Pu fissile nuclides. The development of nuclear energy needs visibility on the demands of the open market and on safety rules, which cannot be anticipated a long time ago due to the complexity of the economy and changing feeling of the society with regard to “nuclear”. The present “nuclear system” (type of reactors and associated facilities) has the objective to produce energy at the lowest cost.

A substantial increase in nuclear energy demand, would not be possible based only on these reactors, because of a lack of resources in Unat and the long time to have enough Pu to launch the necessary FNR to burn U isotopes, reprocessing being mandatory. The use of MOX like today would not be sufficient in this case. The future “nuclear systems” should have the objective to valorise the resources in fissile nuclides and to optimise the management of wastes (including Generation IV reactors). In this section the phase out of nuclear energy is not considered. Such a choice is not so simple to implement, contrary to usual thinking, mainly because stockpile of separated Pu and present wastes (SF or wastes from reprocessing) would have to be managed without the help of Generation IV reactors.

2.1 - Release of effluents, possible dissemination of radioactive matter

The radioactivity of the gas or the fluids (called ultimate effluents), which are released in atmosphere and hydrosphere during operating reactors or facilities of the fuel cycle, are controlled according to the safety rules for radiation protection. The authorisations of releases are calculated on the basis of the dose for the most exposed people at the limit of the site according to scenarios. The true releases are only several per cent of the authorisations. There are also limitations based on the calculated maximum admissible limits in Bq fixed by international organisations for waters, air and some bio-indicators. They derive from scenarios taking into account expositions due to all possible radioactive releases and corresponding to a committed dose (see later) of 1 mSv/year. For instance the limit for T in surface and underground waters is 7 800 Bq/L and in vegetable and fruit the limit is 10 000 Bq/kg or 3250 15

Bq/L of milk. The release of effluents give values much less than these limits. Typical values are as follows. For a NPP (Nuclear Power plant) of four 1300 MWe PWR the yearly release authorised in atmosphere is 3 000 TBq of Kr and Xe, and 10 TBq for other volatile FP (T, I131 and halogens). Only few percents (0.5 to 5 %) are released according to measurements. Release of activity as liquid effluent are 45 to 90 MBq of T and less for others beta emitters. For reprocessing plant it is the same the true measured release are much less than those authorised (for instance 40 GBq of alpha for 1700 GBq, 12 000 TBq of T for 40 000 TBq, 25 TBq of beta-gamma for 1 700 TBq, 1.5 TBq of Sr90 and Cs137 for 220 TBq in liquid effluents and 0,1 GBq of alpha for 75 GBq, 60 TBq of T for 2200 TBq, 5,5 GBq of beta- gamma for 110 GBq, 24 500 TBq of Kr, Xe for 480 000 TBq in gaseous effluents).

The heat transferred to sea or river or atmosphere depends on the power of the reactors both in operation and when stopped (residual power typical of nuclear energy), but such transfers exist for electric plants fuelled with coal or fuel. The release of chemicals by reactors and facilities of the fuel cycle have no special status.

Massive dissemination of radioactive matter (FP and actinides) in environment have occurred due to accidents (Chernobyl, 1018 Bq, Windscale, 7.41014 Bq, Cheliabinsk, 1016? Bq ) or to voluntary release (Techa river, 80 106 m3 of LLLW and HLLLW from reprocessing). There are large areas contaminated around Chernobyl NPP (3700 Bq Pu/m2), old reprocessing plants or facilities of the fuel cycle in the former URSS but also in the USA.

2.2 - Radiological risks (and other risks) potential and residual.

Ionising radiation emitted by radionuclides carry out energy, which can be deposited in living material. The seriousness and the probability of the apparition of the effects produced on health depend on the quantity of energy deposited. The energy lost in matter is measured in Gray (1 G = 1 Jkg-1) and it is called the absorbed dose, D. It is estimated in an organ or a tissue with dosimeters simulating living matter.

The hazard for strong doses (over 1 G) are somatic effects with a probability of apparition of 1 above a threshold and 0 below. The seriousness depends of the dose and can lead to death. Consequently the somatic effects are of deterministic nature. The risk depends of the probability of the occurrence of the event, which leads to the dose. Dramatic examples are known (Chernobyl, use of radioactive sources or beam of particle accelerator). Since the beginning of civilian nuclear energy 54 deaths due to strong doses are recognised. If serious accidents are scarce exposition to low dose are frequent.

In the case of low dose given at low rate the effect is of a stochastic nature : there is a given probability that cancers and hereditary effects appear in a non determined period of time. Here the seriousness of effects is constant, only their probability of apparition depends on the dose. IRPC has chosen a probability of apparition proportional to the dose. To evaluate the effects of low doses on health the absorbed dose in an organ or tissue is not well adapted because each radiation has a different effect on each organ or tissue and a local exposure is different from a global exposure.

One introduces for a given radiation, R, and an organ or tissue, T, the concept of “equivalent dose to an organ or tissue” as: ERT = WR DRT (DRT is the absorbed dose, WR is the factor of nocivity of R for T). Extended to all organs and tissues and all radiation the concept leads to the “effective dose” to the body: E = TWTR WR DRT (WT is the factor of sensibility of T for a 16

given ERT). Due to these corrections the name of the unity the Gray (for DRT), a physical unit, is changed to a biological unit, the Sievert (Sv), still in Jkg-1. E(Sv) is always calculated. For exposition to an external radioactive source there is no problem. Tables give the special -1 -1 values of E for 1 Bq : ER(SvBq ) for any radionuclide and consequently E = R ER(SvBq ) AR(Bq). When exposure to radiation comes from incorporated radionuclides (ingestion or -1 inhalation), and according to bio-kinetic models, Tables give ER(SvBq ), called in this case FDR (dose factor), which correspond to the effective dose given by the radionuclides during -1 50 years after incorporation. Therefore E = R FDR(SvBq ) AR(Bq), and a special name is given to that dose, the “committed effective dose” or in short the “committed dose”. There are specific values of ER for children

Whatever is the dose E: absorbed, effective or committed the radiation risk, RR, associated to -1 -1 -1 -1 a dose is RR (t ) = p (Sv ) i Pi (t ) Ei (Sv ). In this relationship, p is the probability of the -1 -1 -1 effect (p = 1 G or 0 G for strong dose, p = 0.073 Sv for low dose), Pi is the probability of the occurrence of an event i, and Ei is the calculated dose given by the event i. For low dose the unit of time is the year. The basic hypothesis supporting the RR for low doses is called the LNT (linear no threshold). The value of 0.073 (0.06 for a cancer leading to death and 0. 013 for hereditary effect according to IRCP 60) means that 6 cancers per year can appear in a population of 100 000 inhabitants each having got 1mSv. The value 0.073 is on debate and could be reduced to 0.065 (no heredity effect). There are also debates on LNT and the mechanisms of induced cancer by ionising radiations. Notice that as “p” is already a risk RR is the “risk of the risk» for low doses.

-1 -1 IRCP considers that an “acceptable risk” correspond to i Pi (t ) Ei (Sv ) = 20 mSv/year as dose added to the natural dose for a nuclear workers (16 to 65 years old) and 1 mSv/year for any individual of the public. These doses are considered to have any effect on the “expected life”. The natural radioactivity leads to a dose of some mSv/year, 2.4 mSv in average, but fluctuates in large limits depending on local conditions (10 to 50 time the average value). It is clear that according to the LNT law, giving the number of effects as a function of dose, any added dose has an effect and “zero risk” does not exist. IRCP considers that a trivial “added dose” is 1 microSievert (1/100 of the natural dose).

IRCP has a unique definition of an acceptable risk but the perception of risk is another thing. Each one has it own’s. When the scientific knowledge of an event is poor the objectivity of the risk loss its signification and the perception of the risk is preponderant.

For a given amount of radioactive matter (1 ton of SF for instance or quantity to produce

1TWhe) the “inventory of radiotoxicity”, in short the” radiotoxicity”, is defined as i RRi calculated supposing that all radionuclides, i, are incorporated. It represents a potential risk.

The residual risk is i RRi calculated for a given case of exposure . It takes into account the management of the radioactive matter(like geological disposal for instance).

Doses due to the use of nuclear energy are low compared to the natural doses and those given by medical diagnostic (1.5 mSv/year in average per person)

2.3 - Impacts in relation with sustainable development

Technological criteria 17

At least 4 technical criteria to assure sustainable development have to be fulfilled by nuclear energy (and any source of energy) : providing adapted power to the energy needs, safe technology, durability of resources (Unat), no collateral damage (proliferation of nuclear material). Management of nuclear wastes for the coming years rise some technical problems but the main problems are of societal type.

Nuclear energy can only contribute to satisfy, presently, the need of electricity. As already said in most nuclear countries it supplies conventional sources to an amount of 30 %, except France where almost all electricity comes from nuclear energy (80-85%) and dams. The world-wide tendencies are discussed above. As also discussed above, nuclear energy can be considered as safe. The next Generation III reactors, like EPR, will have improved systems to reinforce safety. The resources of Unat are secure for the next decades at known prices. Ultimate resources are estimated to 17 Mt but at an unknown price. Pu can be recycled once in PWR without difficulty. Contribution of fuel to the price of kWhe is around 30 %.

The proliferation is a question of technical and political means controls through NPT under the auspices of ONU and AIEA. NPT is the Non Proliferation Treatise on nuclear weapons ratified by 190 countries which have already nuclear weapons or do not want to have. It is easier to get fissile nuclides from high level enriched Unat in U235 (ultra-centrifugation) than from burning Unat to few MWjt-1 in thermal neutrons nuclear reactors to produce Pu239. Both techniques are very heavy. Fuel for research reactors is enriched U (at least 20 %). As already said the isotopic composition of Pu issued from SF is not the best to prepare a nuclear engines. “Dirty bomb” could be prepared with radionuclides used in nuclear technology. Of course diversion of radioactive matter cannot be avoided.

Societal criteria

With regard to societal criteria there are 2 important points : cost and wastes management, the radiological impact of nuclear energy being low (except in the case of accidents)

Cost of nuclear kWhe has been subject of much estimation since 20 years. It is a complicated question depending on numerous factors and relevant to “case studies”. There is presently a consensus that the true price is around 3 to 4 cents of Euro per kWhe (including provisions for waste management and decommissioning) whatever would be the interest of money (5 to 10 %) on the next 40 years. The nuclear kWhe is and will be cheaper than those of gas, coal and wind. The price of Pu is taken as zero when MOX is used (compensation between reprocessing and economy on Uenr). But this question is under debate. The breakdown of price between investment, exploitation, and fuel depends on the mode of production.

Extraction of Unat from ores (when the content is above 0.2-0.5 %) leaves an enormous amount of wastes (5 107 tons in France) which contain Th230 and its daughters (Ra226 and Rn222) but with a low activity (107 Bq/t). Their management requires essentially to confine Rn. “Short-lived wastes” are disposed of in ground or sub-surface facilities (or stored waiting for a definitive issue). Management of HLLLW and MLLLW are the main concern of people. The dimension of the problems raised by HLLLW and MLLLW can be appreciated looking at the quantities to be managed and the periods of time to be considered.

Actually 10 000 tons of SF are yearly unloaded in the world. Those reprocessed give wastes packages, Unat and some separated Pu. Nobody knows exactly if it will be possible to transmute on an industrial scale the long-lived radionuclides and which ultimate waste would 18 be issued by the new reprocessing methods. That could take, in an optimistic view, several decades in the framework of the continuation of the use of nuclear energy over, say, two centuries. The other options are indefinite storage by multiple rebuilding of present storage facilities and a deep geologic disposal. A geologic repository could be designed to dispose of 80 000 tons of SF packages (including 800 tons of Pu) or more or less the equivalent wastes from reprocessing of these SF, but without U and Pu (or relatively low quantities), which would have to be managed according another way. In 2020 , 200 000 tons of SF will have to be managed. If geological disposal is chosen that will need 3 disposal sites. An increase of the use of nuclear energy must be considered with respect to the need of more disposal sites (10 time more will lead to one site a year!). Transmutation of long-lived FP and actinides is based on 3 criteria: kinetics must be compatible with fuel cycle management and half-life of radionuclides, quantities loaded must be compatible with safety reactor and larger than that produced in the same period of time, build up of undesirable nuclides must be avoided as possible. Accordingly, among FP only Tc99 and I129 fulfil the criteria. Actinides nuclides which have high radiotoxicity (FD around 2 10-7 SvBq-1) could be transmuted (Pu and Am are the most radiotoxic actinides and must be burnt in priority).

There are 4 characteristic periods of time in long-lived nuclear waste management. The first extend to several decades (5 to 10 ?) during which SF or HLLLW must be cooled, due to the heat released by FP and short-lived actinides (Pu241, Cm244, Cm243 for UOX and additionally Pu238 for MOX). During this time either the way to change the type of wastes or to dispose of the wastes is decided. The second period may span also over several decades to implement the choice. There is no need of cooling the waste packages during this period. The problems during these two first periods are of national relevance and solved by man-made technique. In the case of geological disposal heat released by Am241 and Pu238 lay down the size of the installations. The third period extends to 100 000 years during which radionuclides must be isolated/confined, for instance in canisters and engineered barriers of high performances (Pu239, Pu240, Pu242, Am241 and Np237). Over 100 000 years the radionuclides (U isotopes, Np237 and Pu242 and also long-lived FP) must be confined by natural rocks. At very long term one can shows that U235 = U235 in + Pu239 in, U236 =

U236in + Pu240 in, U238 = U238in +  Pu242in, Np237 = Np237in + (Pu241in and Am241in),

Pu242 = (1 - ) Pu242in. During these two last periods the problems can only be solved by “geology” and the concerns could have a world-wide dimension (like CO2) because in a large number of generations every people could have a chance to be near outlets of disposals.

3 - Is nuclear energy renewable?

The question: is nuclear energy a renewable form of energy? is a matter for the next half century. Nuclear energy is not renewable like solar or wind energy because Unat is an earth natural resource like coal or petroleum. The present yield for the utilisation of the energy of fission contained in Unat is less than 1%. But when U238 is used as fissile and fertile isotope the period of time during which energy can be produced is measured, as least in theory, in thousands of years. In the case of lack of U, or in parallel for technical reasons, the use of Th can also be considered on the same scale. But the use of new reactors and new “nuclear systems” is mandatory.

3.1 - Reserve of other “uraniums” than Unat, further demand on Unat

The stockpile of Udep is enormous (actually around 230 000 tons in France for instance) and will increase with the use of nuclear energy based on U235. Udep is easy to manipulate. As 19 seen before the quantities of Uret are less (actually around 20 000 tons in France) and will also increase. Due to the presence of U236, U232 (coming from Pu236 and daughters) and traces of FP or actinides it cannot be enriched so easily than Unat. So its use is restricted today to prepare some sub-assemblies of MOX. Declassified military high enriched U (up to 90 % in U235) could be used. World-wide quantities of all of these kinds of U can be estimated according to average figures concerning methods of enrichment (5 to 9 tons of Unat to get 1 ton), total unloaded SF (250 000 to 300 000 tons), total SF reprocessed (75 000 tons) and declassified weapons.

According to some forecast the need of Unat could be important. For instance a power of 250 GWe in 2020 would require 100 kt per year of Unat. So extraction of Unat both from pure U ores and as by-product of extraction of Cu, Au or P should be boosted. Recent discovered U ores could be as rich as 15 % or more (80 % in Cigar Lake site, Atabaska, Canada). New or improved methods of extraction are under investigation. Prospecting of U under 1000 m in earth crust are considered.

3.2 - Reactor for the future (2050), valorisation of resources, optimisation of waste management.

The nuclear energy for the future will be developed in the direction of mixing this source of energy with other sources in an “energetic mix” whatever the other objectives are. A distinction have to be made between “ systems of reactors” and “nuclear systems of energy” which can includes several systems of reactors because a given reactor, even new , as achieved as possible, cannot burn a maximum of Pu, use the maximum of fertile nuclides and minimise long-lived radionuclides production. One has to consider systems of reactors, which can valorise U (use of PWR, BWR, CANDU and FNR), burn Pu (PWR and HTR), burn actinides (RNR or ADS), produce Pu (RNR) and use Th. The perspective to reach a given “nuclear system of energy” by mixing “systems of reactors” at an industrial level is unknown.

The most advanced new reactors are the following.

HTR (High Temperature Reactors). Two international projects of HTR of low power (100 and 300 MWe) are developed. The technology derives from experimental reactors, which have been operated in USA and Germany in the sixties-seventies. Presently there are two experimental HTR (10 MWth in China and 30 MWth in Japan). The coming reactors will be fuelled with Uenr (8 to 10 %) or with MOX made with military Pu239, moderated with C, cooled with He and operated following an open cycle (75 % of loaded Pu could be burnt). With He at 600°C directly associated with a gas turbine a yield of 50 % is expected (Brayton cycle). Their fuel will be based on micro-spheres of ceramic oxides (or carbides) coated by several layers of C and by SiC and embedded in C to give pebbles (PBMR- Pebble Bed Modular Reactor project) or prismatic fuel (GT MHR project). These layers will isolate FP and actinides from He (like cladding of pins in PWR) up to 1600 °C. SF at high BU (100 to 150 GWjt-1) will be a waste because its reprocessing will be very difficult (but not impossible). Power of HTR is limited to allow a passive cooling in case of leak of He and of stopping of the reactor. HTR have electric power and fuel cycle adapted to developing countries. They open the perspective to produce H2. They could be developed to VHTR (Very High Temperature Reactor, 1000 °C) and the use of fast neutrons.

FNR. Fast neutrons allow the use of all isotopes of U, Pu and heavier actinides (f/c of fast neutrons is higher than f/c of thermal neutrons). The resource of fissile nuclides is 20 practically increased by a factor of 50 (twice in theory). The technology of FNR cooled by Na is known. They are fuelled with MOX of high content in Pu (up to 20 %) and need to be launched 10 to 15 tons of Pu/GWe (in fact with 9.6 tons of fissile Pu isotopes with the equivalency factors as follows : 0.81 for Pu238, 1 for Pu239, 0.35 for Pu240, 1.05 for Pu 241, 0 for Pu242). They can be operated giving as much Pu as they burn (regeneration) or more (over-generation or breeding). But it takes 2 to 3 decades to have sufficient Pu to launch a new FNR (50 years for a PWR). Typical figures for Superphenix (1.2 GWe, now in decommissioning process) were 5 tons of Pu in the core, expected 0.8 tons burnt and 0.96 build up in blanket per year (30 years to get 5 tons of Pu). Worldwide Pu production is around 75 tons (and 7.5 tons of other actinides), which could allow to launch 5 to 7 GWe/year. In 2030 the stockpile of Pu in SF will be around 3000 tons. One can make previsions based on these figures but investments are enormous. The possible development of FNR incite to look at Pu as a strategic resource and to store it as SF UOX and MOX. It leads also to improve reprocessing methods. Use of Na raise problems due to chemical reactivity (air and water), k/k > 0, safety inspection of core vessel, which are not encountered when coolant is a gas. VHTR FNR He gas cooled are seen as reactors in line with FNR Na cooled and HTR but new fuels must be developed. Fuel in nuclear energy is always a key problem.

ADS (Accelerator driven system). The ADS are based, like FNR, on fission induced by fast neutrons. But the core (U) is sub-critical (keff around 0.98 for instance) and the fast neutrons needed to have  = 0 are given by a spallation source. In this device (cooled molten Pb-Bi alloy, diameter 0.5 m, height 1 m) a beam of high energetic and high flux of proton (1 GeV, 20 mA) is transformed in fast neutrons (1 to 2 MeV) by nuclear spallation reactions. Around 33 MeV are needed to produce one fission (the yield is around 6). An accelerator of high performances (at the limits of the present technology) is needed to accelerate the protons, which have to pass a window before hunting the spallation source. Parts of ADS (core, source, accelerator) are on experimental evaluation before to decide on a prototype. Here the power of the reactor, Pr, is controlled by the power of the accelerator, Pa. The ration Pr/Pa =

6 / (keff/1-keff). With keff = 0.98,  = 2.5 and  = 1 (factor depending on the geometry core- source) to have 1 GWe or 2.5 GWth needs a power of 20 MW (1GeV, 20 mA). The use of ADS is foreseen to transmute Am (and/or Pu) embedded in inert target (without U). Transmutation with ADS allows a load in actinide higher than in critical FNR because operating the reactor dos not need delayed neutrons (which are scarce with heavier fissile nuclides than U235).

Nobody knows the price of the kWhe of these new “nuclear systems”. Estimation are 1.25 to 1.5 time that of PWR.

3.4 - Other reserves of fertile nuclides

It is possible to launch thermal neutrons based reactors using the fertile monoisotopic Th232 and fissile nuclides (U235 or Pu isotopes for instance). Fissile U233 is formed, which can be recycled. U233 has very attractive properties ( and  % higher than in the case of U235 and

Pu239). Fission induced by thermal neutrons is important (f/c = 9) contrary to Pu239 (f/c = 2.3). U233 can be, or could be, produced in PWR by irradiation of Th232. Some prototypes of Th reactors have been operated with more or less success during 20 years since 1970 (HTR with ThO2 in USA and Germany) and India experiment a “Th cycle”. A molten salts reactor (MSR) in USA (7.5 MWth, U235 and U233 fuelled) has shown, in the sixties, the possibility of breeding. A modern version is under evaluation. MSR are high temperature reactors (600 to 700 °C) where molten salts (Li and Be fluorides) are both fuel and coolant and the 21 moderator is C. It needs 1.2 tons of fissile nuclide per GWe, around 1/10 of a FNR-Na load. It does not produce heavy actinides (U238 is not present) or in low quantities: Np237 0.59 kg/TWhe, Pu238 (0.14) Pa231 (0.23). FP have to be extracted from the fuel periodically or in line (pyrochemical processes) with the results that neutrons flux is constant. Such reactor can transmute actinide in line (20 kg/TWhe in theory). Many problems have to be solved at each step of the “Th system”. Liquid fuel is attractive but raises new problems: containment of radioactvity (like T coming from ternary fission and (n,p) nuclear reactions on Li6 ), cooling for residual power, corrosion, radiological protection against 2.6 MeV  from Tl208, a daughter of U232 produced by U233(n,2n)U232 and new wastes (containing U232, U233, U234, Pa231, Np237). Therefore waste management would be only governed by FP. The Th system has balanced advantages/disadvantages and its development is not foreseen unless massive fertile nuclides would be necessary, if so.

4 - Conclusion

Technology and economy make fission nuclear energy a “sustainable energy”: Unat resources for decades (and possible use of fertile nuclides), safe technology and possible improvement, low price of kWhe compared to other sources (without considering CO2 emission tax), environment friendless and no health impact by additional low dose of radiation. A renew of nuclear energy could be possible. But problems remain: public antipathy (difficult to change), waste management (problems identified, but not solved), international policy for licensing (visibility for development), ethical (intergeneration relationships).

It is reasonable to forecast that during the next 15-20 years (say up to 2025) no drastic change will occur in nuclear energy production. This period will be for each country a period of thinking on the use of nuclear energy, confirmation of phase out, maintaining present level or increasing it in the “energetic mix”. For countries which will confirm the nuclear energy option the first objective will be the prolongation of the life of the reactors and to replace those no longer allowed to be exploited by the safety authority. Reactors of “Generation III” are ready to be launched. The demand of Unat will depend of options but according to engaged programmes it will slightly increase (70 000 tons in 2005 to 85 000 in 2015). Few countries reprocess their SF and except a total change in USA policy the situation will remain as it is. Basically the production of nuclear wastes (SF or glass packages) will stay at the present level or will slightly decrease due to improvements in BU of fuels and conditioning. One can think that during the next 20 years actions to manage nuclear wastes will progress, particularly considering geological disposal, which, finally, appears difficult to be bypassed whatever will be the “ultimate wastes”. Finland has decided to dispose of SF in an accepted site (in operation in 2020 ?). This period will be also a test period for the implementation of programmes set up to renew world-wide nuclear energy as for instance the GNEP (Global Nuclear Energy Partnerships) leaded by USA. Finally it will be devoted to test the will of countries in developing international research to prepare possible launching of reactors of “Generation IV”, and associated nuclear fuel cycles, in the second half of the century according to the objectives of GIF.

GIF (Generation IV International Forum, 11 partners) was initiated in 2002. Several organisational steps have been implemented up to 2005. It is aimed at developing 6 new types of reactors based on fast and thermal neutrons for optimising the use of fertile nuclides, producing less waste and opening new uses of nuclear energy (high temperature heat production). The GNEP organisational programme, launched in 2005 by USA, proposes to complete the objectives of GIF as follows. In the short term: encourage launching of new 22 reactors particularly for developing countries (low power), in the long term: develop new technologies, proliferation resistant, for recycling Pu and other actinides and develop advanced burners of Pu and actinides. For both actions it is proposed to set up an international system of nuclear services (enrichment, reprocessing) under international control

What will happen after 2025 for nuclear energy is relevant to prospective because choices on energy are subject to too many parameters. In the countries where nuclear option will remain open one can think that coexistence of Generation II and III of reactors will exist. As the Generation III reactors are designed for a 60 years lifetime these reactors, and associated fuel cycle facilities, will finally dominate the nuclear landscape up to the end of the century. This will not lead to a great change in nuclear industry. In the case of positive tests for a possible development of nuclear energy, which will mean that a drastic increase in nuclear energy will have been accepted, research for “nuclear energy for the future” will be boosted to prepare the use of fertile nuclides (U238 and Th232) and to implement the objectives of GIF.

Load more

  22

Copy Link