Phase diagram of water

Thermal Power
Origins of Steam Power
Papin (1690)
First Steam-operated piston
i) water boiled in cylindrical chamber
containing tight fitting piston
ii) steam exerted force on piston,
causing it to rise
iii) piston retracted into chamber
after walls cooled down
(cycle time – many minutes)
Savery (1699)
Rate of steam condensation increased by spraying cold water over outside of piston
chamber (cycle time ~ a minute)
Newcomen (1712)
Rate of steam condensation improved still further by injecting cold water directly into
steam chamber (cycle time 5-10 seconds)
Watt (1775)
Incorporated separate condenser, thereby removing need to reheat walls of piston
chamber. Commercial steam engines of 20 kW power in use by 1800
Thermal Power Stations
Note: thermal includes fossil-fuel and nuclear power
Heat source is part of Steam Cycle
Thermodynamics of cycle independent of nature of heat source
Steam Cycle: Main Components
Water
Pump
Cooling water
Heat in
Condenser
Heat out
Boiler
Turbine (expander)
ω
Electrical power
Phase diagram of water
Temperature K
T
Super-critical
fluid
Properties of Steam
Critical pressure 221 bar
Sub-critical
dry steam
p=0.006 bar
Sub-critical
water-steam
mixture
(wet steam)
100% water
0% steam
0% water
100% steam
Ice-water vapour mixture
s
Specific entropy kJ/kg/K
Dryness fraction (quality)
x = m vapour
/(m vapour
+m water
)
s = (1 − x)s water
+ xs vapour
Entropy/temperature diagram is best for power station cycles
Any TWO thermodynamic parameters are sufficient to define state of fluid
eg S,T or P,H (Steam Tables)

Carnot Cycle (Ideal Cycle)
T
Q 12
T 1 2
a
W 41
W 23
Q 34
T b
4 3
S
1) Heat absorption at constant
temperature, T a (boiler) 12
2) Isentropic expansion work
output (turbine) 23
3) Heat rejection at constant
temperature, T b (condenser) 34
4) Isentropic compression
(pump) 41
Energy Conservation (1 st Law of Thermodynamics)
Q 12 + W 23 + Q 34 + W 41 = 0
(Note: Q 12
> 0, W 23
< 0, Q 34
< 0, W 41
> 0)
Cycle efficiency, η c = (Useful work out)/(Heat input at T a )
ie η c = (| W 23 | − W 41 )/ Q 12 | = (T a − T b )/ T a = 1 − T b / T a
(Note: T measured in K (absolute temperature) – formal
definition of absolute temperature scale)
Practical difficulties in using a Carnot Cycle
1) Boiler operates only in wet-steam regime otherwise temperature
would rise when all the water has turned to steam, violating
condition for Carnot Cycle
turbine expands wet steam
water droplets hit turbine blades (damage)
2) Maximum temperature (T a ) is limited to ~650 K
efficiency of cycle is severely constrained
3) Compression of water/steam mixture is thermodynamically
unstable (water droplets)
very large volume compressor (expensive)
Rankine Cycle overcomes all these problems
T
2b
2a
T b
1b 1a
2c
S
Rankine Cycle
Step 1:
a) Condense all the steam to water
in the condenser
b) Pumping water to high pressure requires
small volume machine and little energy
T
T b
T a
S
Step 3:
Expand dry steam through a turbine
to generate shaft power
In practice, water droplets still form in
the low pressure end of the turbine, so
the steam is reheated at various stages
Step 2:
Use 3-stage boiler (~ constant pressure)
a) Economiser – water heated at constant pressure
b) Evaporator – water/steam mixture heated at constant pressure
c) Superheater – dry steam heated at constant pressure
[Note that there is a small drop in pressure through the boiler tube
in order to overcome frictional losses]
T
T a
T b
S
HP
from
IP LP
boiler
reheaters
HP: high pressure turbine
IP: intermediate pressure turbine
LP: low pressure turbine
to condenser

Frictional losses across turbine blades vary like u 2 (F D =½C D ρAu 2 )
ie very large for large u (near speed of sound)
Losses reduced significantly by using many stages in series (~50 stages)
Other practical effects limiting efficiency
The loss of kinetic energy at each stage is
small and turbulence is reduced
a) Boiler tubes have finite thickness, so outer wall temperature is higher
than water/steam temperature
b) Metallurgical limit to temperature/pressure difference boiler tubes
can withstand (creep/crack formation)
c) Many pipes/tubes in flow circuit frictional losses
d) Condenser is a vacuum chamber air leaks in but can not condense,
so ‘air blanket’ forms, preventing water vapour from condensing on
cold surface of condenser tubes
T
W 12
2
1
3
Q 23
W 34
Q 41
4
Efficiency of Rankine Cycle
S
Condenser at 30 C at a pressure of 0.04 bar
Compressor increases pressure to 170 bar
Three-stage boiler at 170 bar
a) economiser raises temperature to 352 C
b) evaporator at 352 C
c) superheater raises temperature to 600 C
Adiabatic turbine
T p h f h g s f s g
Water/Steam 30 0.04 126 2566 0.436 8.452
Water/Steam 352 170 1690 2548 3.808 5.181
Dry Steam 600 170 3564 6.603
where h f and h g are the specific enthalpies and s f and s g are the
specific entropies of the fluid and gas, respectively, in kJ/kg.
Adiabatic compression or expansion
W = W s + (p 1 v 1 − p 2 v 2 ) = ∆U = u 2 − u 1
W s = (u 2 − u 1 ) − (p 1 v 1 − p 2 v 2 )
= (u 2 + p 2 v 2 ) − (u 1 + p 1 v 1 )
= h 2 − h 1
Work done by the shaft W s
on the fluid
Adiabatic so Q = 0
Total work W
First Law: ∆U = Q + W
Note sign convention
In adiabatic process work done equals change in enthalpy
Specific enthalpy h = u +pv; dh = TdS + Vdp
isobaric, constant pressure, dh = du + pdv = dQ
isentropic dh = dW = Vdp
i) W 12 = V(p 2 − p 1 ) = 10 −3 (170 − 0.04) 10 5
=17 kJ/kg 3
ii) 12 isentropic so
h 2 = h 1 + W 12 = 126 + 17 = 143 kJ/kg
iii) 23 isobaric so
Q 23 = h 3 – h 2 = 3564 − 143 = 3421 kJ/kg
iv) 34 isentropic so
W 34 = h 3 – h 4 and s 3 = s 4
s 4 = (1−x)s f4 + xs g4
s 3 = 6.603 = (1−x)0.436 + 8.452 x
x = 0.769
T
W 12
2
1
Q 23
W 34
Q 41
4
S

Combined Cycle Gas Turbine (CCGT) Stations
v) h 4 = (1−x)h f4 + xh g4
h 4 = (1−x)126 + 2566 x
x = 0.769
h 4 = 2002 kJ/kg
In recent years gas turbines and steam turbines have been combined to
increase the efficiency to around 50-60% (upper temperature ~1200 C)
Gas Turbine
air
34 isentropic so
W 34 = h 3 – h 4
= 3564 – 2002 = 1562 kJ/kg
vi) η = useful work/heat in
= (W 34 – W 12 )/Q 23 = (1562 – 17)/3421
= 0.452 = 45.2%
vii) cf Carnot Cycle
η c = (T 3 − T 4 )/T 3 = (873 − 303)/873
= 0.653 = 65.3%
T
W 12
Q
3
23
2
W 34
1
4
Q 41 S
T
T max
Compressor
Combustion
Compressor
Gaseous
fuel
p
compressed
air
Combustion
Chamber
p atmos
Turbine
Brayton Cycle
S
Turbine
Exhaust gas
a) Heat generated by internal combustion rather than
via a high temperature heat exchanger (boiler)
b) No cooler required since exhaust gases vented to
atmosphere
Plant much smaller. Work done by compressor is
significant, though this is compensated by very high
temperature ~ 1200 C (Turbine blades ceramic coated
and water cooled)
ω
CCGT Station
660 MW Power Plant
air
Compressor
Gaseous
fuel
compressed
air
Combustion
Chamber
Turbine
Exhaust gas
Water
Pump
Heat in
Boiler
Turbine
Exhaust gas
Condenser
ω
combustion
Cooling water
Heat out
T
Rankine
Cycle
compressor
S
turbine
boiler
Brayton
Cycle
Heat of exhaust gases used to
raise steam for steam turbine
Many CCGTs have been built in the
UK in the 90s due to availability of
cheap gas and relaxation of
governmental controls
Stator for a 660 MW
generator being assembled
Low pressure turbine, part of a
660 MW assembly
Power density ~ 1000 MW e per sq km

Types of Fossil Fuel Power Stations
CCS
Type
T o C
bar
Efficiency
Subcritical
538
167
up to 39%
Supercritical
540-566
250
up to 46%
Ultra-supercritical
580-620
270-285
50-55%
IGCC: Integrated gasification combined cycle
Coal + O 2 + H 2 O → H 2 + CO (syngas)
mainly Power Plant Today Future
Syngas used in combined cycle gas turbine (CCGT) power station
Efficiency ~ 45%
Shift reaction : CO + H 2 O → H 2 + CO 2
Pre-combustion Carbon Capture and Storage (CCS)
Conventional Coal 24-40% 15-20%
Natural Gas 11-24% 8-11%
Advanced Coal 14-25% 9-12%
Typical Energy Penalty(increase fuel use per KWh produced due to CO 2 capture)
Potential by 2050
~1000 GW e with CCS
Power density
~ 1000 MW e per sq km
CCS
CCS
Technical Solutions to Disposing of CO 2
• Underground storage
In aquifiers, used gas/oil fields - huge storage potential, but
possibility of spontaneous gas eruptions (1750 people killed by CO 2
eruption from volcanic lake in 1986)
•Deep ocean disposal
Large hydrostatic pressure CO 2 liquifies
- long-term viability uncertain
- effect on ocean deep-sea creatures uncertain (affects food chain of
surface creatures
Air Capture
Electrocatalytic CO 2
Conversion to Oxalate
by a Copper Complex.
SCIENCE VOL 327 15 JAN. 2010
• Pump CO 2 into lakes/breed algae
Algae dried biomass alternative to fossil fuel
Or
Algae biodiesel alternative to petrol or diesel from oil
(CO 2 –neutral if re-used; sequestered if buried)

Historical Milestones
Nuclear Power
1896 Becquerel Fogging of photographic plates near U salts
1905 Einstein Special theory of relativity- E = mc 2
1911 Rutherford Discovery of nucleus - α-particle scattering
1913 Bohr Quantum model of H atom
1932 Chadwick Discovery of neutron
1936 Bohr, Frenkel Liquid drop model of nucleus
1938 Hahn, Strassmann Discovery of fission
1939 Joliot, von Halban Discovery of neutrons produced in fission
Kowarski reactions possibility of ‘chain reaction’
1939 Szilard, Wigner Advised Roosevelt of feasibility of uranium
bomb
1939 Booth, Dunning, Start of projects to separate isotopes of
Urey
235
U and 238 U
1940 Anderson, Fermi Showed that 12 C would be a good moderator
1940 Joliot, Dautry Transferred D 2 O from Norway to UK
1940 Seaborg Discovered Plutonium
1942 Groves Manhattan Project started
1942 Fermi First nuclear reactor- demonstrated that
chain reaction controllable
1943 Bethe, Weisskopf Defined specification of atomic bomb
Teller, Feynman (sub-critical sphere surrounded by
explosives compression criticality)
May 1945
July 1945
Aug 1945
Experimental uranium bomb exploded
Experimental plutonium bomb exploded
Hiroshima destroyed by U-bomb
Nagasaki destroyed by Pu-bomb
First self-sustaining chain reaction
Fermi December 1942 Chicago University Stadium
1945 UKAEA established in UK, CEA established in France
1954 Fast reactor programme started
1956 First prototype power station (Calder Hall) – gas cooled
1956 Suez crisis oil shortage nuclear power stations
first commercial reactors 1962 (Berkeley, Bradwell)
1957 Pressurised Water Reactor (PWR) developed for
nuclear submarines by the USA
1957 Windscale fire (Wigner energy underestimated)
1957 Campaign for Nuclear Disarmament (CND) established
1959 Dounreay fast reactor critical
1964 UK decide to build Advanced Gas cooled Reactors (AGR)
1976 First AGRs commissioned (Hinckley B, Hunterston B)
1979 Three Mile Island accident (operator errors)
1986 Chernobyl accident (design faults, operator errors,
no regulation
1991/2 Collapse of communism in E.Europe nuclear
cooperation (civil and military)
1995 First PWR in UK (Sizewell B)

B/A
Fusion
Binding Energy of Nuclei
Fission
Mass Number A
MeV
Above mass ~20 approximately constant binding energy per nucleon
However: more stable nuclei can be formed either by:
i) Fusion (combining 2 nuclei with low mass number A)
ii) Fission (breakup of large A nucleus into lower A fragments plus
release of neutrons)
8
6
4
2
0
In fission
A 1 → A 2 + A 3 + neutrons,
where A 2 and A 3 are the final
stable nuclei, the total energy
release E R is approximately
E R = A 2 {b(A 2 ) − b(A 1 )} +
A 3 {b(A 3 ) − b(A 1 )}.
Basic Ideas : Fission Reactors
141
Ba 56
n
n
235 236
U U n
92 92
n
neutron
92
Kr
absorption
36
fission
n + 235 U 92 141 Ba 56 + 92 Kr 36 + 3n
Change in mass, δm = 3.6 10 −28 kg
Energy released, E = (δm)c 2
= (3 10 8 ) 2 3.6 10 −28
= 3.2 10 −11 J
⇒ chain reaction
cf chemical combustion
C + O 2 CO 2 E = 7 10 −19 J
Energy release from 1 uranium nucleus = 5 10 7 carbon atoms
1 tonne of 235 U = 2.7 10 6 tonnes of coal
U is 0.7% 235 U so 1 tonne U ≡ 20,000 tonnes of coal
Naturally-occurring uranium consists of
• fissile isotope 235 U
• stable isotope 238 U }ratio = 1/138 ~ 0.7%
In a reactor, neutrons are lost by :
1) absorption by 238 U 239 U 239 Pu
2) absorption by 235 U 236 U (18% for thermal n)
3) absorption by moderator
4) absorption by reactor structure
5) escape from reactor core
Not enough n to continue chain reaction with H 2 O moderation
enrichment necessary to increase ratio 235 U/ 238 U 3%
Note: the naturally-occurring ratio was higher in the past
%
Oklo ‘reactor’ (Gabon)
3
0.7
−2.10 9
−10 9
0
years
β
nb t 1/2 of
235
U = 7.1 10 8 years
cf t 1/2 of
238
U = 3 10 9 years
Fuel Enrichment
Enrichment is process of increasing proportion of fissionable nuclei
in natural uranium (0.7% 235 U)
Methods of enrichment
1 Electromagnetic Separation
accelerating
electrodes
B
mv 2 /r = qBv r = mv/qB
vacuum chamber
heavy isotope
light isotope
Used for Manhattan project (1g/day) and by Iraq before Gulf War

2 Gaseous Diffusion
Uranium ore converted to UF 6 gas passed through very thin porous
membranes. Light 235 U molecule diffuses faster than the heavier
238
U molecule. ~1400 stages to achieve 3-5% 235 U/ 238 U
Enrichment
3 Ultracentrifuge
Gaseous UF 6 is rotated at high angular velocity in a cascade of
centrifuges, causing partial separation
4 Laser Separation
Tuned lasers selectively ionise the lighter isotope in UF 6 vapour
Positive ion attracted to charged collector plates – still being
developed
Energy Released by Fission Process
Instantaneous Release (per fission) MeV
Fission products
168 heat
Neutrons 5
γ-rays
7 heat
Delayed Release (per fission) MeV
β-particles
8 heat
γ-rays
7 heat
Antineutrinos
12 lost from reactor
Neutron Capture by 238 U
β
β
n + 238 U 92 239 U 92
239
Np 93
239
Pu 94
23.5 m 2.3 d
In practice many neutrons do not contribute to fission because
they are absorbed by 238 U
f(E)
0 2 4 6
Neutron energy (MeV)
Neutron cross-sections on Uranium
Barns
10 3
10 2
10 1
10 0
σ∼1/v
Neutron Energy Distribution
average energy
Neutron energy density, f(E)
f(E) ~ 0.77(E) 1/2 exp(−0.775E)
On average, 2.44 neutrons are
produced per fission
Average neutron energy is ~ 2MeV
fission by 235 U 92
capture by 238 U 92
fission by 238 U 92
10 −1 10 1 10 3 10 5 10 7 eV

Macroscopic cross-sections for natural
uranium (Σ t
= Σn i
σ ti
)
Factors affecting chain reaction
1) For each thermal neutron absorbed, η effective fast neutrons emitted
η < ν, mean number produced (ν =2.42 for 235 U), because not all
neutrons absorbed by fuel cause fission. Nat U (0.72% 235 U) η =1.33
2) Some fast neutrons cause fission before slowing down which
increases the number of neutrons by the fast fission factor ε
3) The probability that a neutron will avoid resonance capture by 238 U
the resonance escape probability p - depends on the moderator
4) The fraction of thermal neutrons that are absorbed by the fuel in the
core (fuel, moderator, can) is called the thermal utilization factor f
5) There are a fraction l f of fast neutrons and a fraction l t of thermal
neutrons that leak out of the reactor
The neutron multiplication factor k is therefore given by:
k = ηεpf(1− l f ) (1− l t )
For infinite core k ∞ = ηεpf
Four factors formula
Neutron Moderation
Moderator is a medium for reducing the kinetic energy of neutrons from
MeV to thermal level without losing many in the ‘resonant trap’ of 238 U
m
M
For 180 deg scattering
neutron
E s = [(M − m)/(M + m)] 2 E i = [(A − 1)/(A + 1)] 2 E i
For 0 deg scattering
E s = E i
Averaging, E s = ½{1 + [(A − 1)/(A + 1)] 2 }E s
nucleus
= [(A 2 + 1)/(A + 1) 2 ]E s
(Averaging over all angles gives the same result)
1
H
12
C
238
U
A 1 12 238
E s /E i 0.5 0.86 0.99
How many collisions required to reduce neutron energy from
2 MeV to 1 eV ? (factor of 2 10 6 )
Put (E s /E i ) n = 1/(2.10 6 ) = 5.10 −7
eg 1 H gives n ~ 21, 12 C gives n ~ 96
Moderating Ratio, MR
Good moderators require
• large σ elastic (σ el )
• low σ capture (σ c )
• significant loss in KE per collision
• chemical stability (in hot, radioactive environment)
Moderating ratio, MR = (1− E s /E i ) σ el /σ c
Reactor Control
H 2 O 62; D 2 O 4830; C 216
If the neutron flux increases to a higher level than that needed for a
stable chain reaction, how can the reactor be controlled, ie how can
equilibrium be restored?

Lower ‘control rods’ into reactor to absorb excess neutrons
Materials used: 113 Cd (σ c = 20,000 barns)
10
B (σ c = 4,000 barns)
cross section is for thermal neutrons (~0.025 eV)
[σ c (max) ~ π(λ/2π) 2 ∼ 2.6 10 7 barns]
(withdraw control rods if reactivity gets too low)
In practice control would be virtually impossible but for the existence
of ‘delayed’ neutrons
Delayed neutrons are released only after the β-decay of a fission product
Typically, about 1% of neutrons produced by fission are delayed by
10-20 seconds, which is enough time for small adjustments in the
position of the control rods (automatically controlled)
eg
87
Br β −
t 1/2
54.5 s
n
87
Kr *
86
Kr
137
I β −
t 1/2
21.8 s
n
137
Xe *
136
Xe
Neutron Population Growth
η is number of neutrons emitted per neutron absorbed
Because of losses the mean number is k, where k < η
k is the effective multiplication factor
If all neutrons were prompt the neutron population would grow like
dn/dt = n(k−1)/τ = nq/τ
where q=(k−1) and τ is the average neutron lifetime in the reactor
So
n = n o exp{qt/τ}
eg q = 0.001, τ = 0.001 second gives n = n o exp(t), so after 10 s
n/n o increases by a factor of 22,000
In 235 U the mean lifetime of the
groups of delayed neutrons is
about τ d = 9 s and they represent
a fraction β = 0.65% of the total
neutron emission
235
U delayed neutrons
If q

Safety Features in a PWR
• The control rods can be lowered fully in the case of an emergency
• Should the pressure drop in the primary loop and the water start to
boil, the creation of bubbles (voids) decreases the moderation and also
the absorption. The effect on the moderation is the more significant
and the chain reaction stops and the reactor is no longer critical
• The moderation is also decreased if the core temperature rises, as
this increases the Doppler broadening of the 238 U resonances, which
decreases the resonance escape probability p
• A loss-of-coolant accident (LOCA) in which the water in the
primary loop is lost requires additional emergency cooling to be
available. The outer containment vessel provides a final barrier and
worked successfully in the Three Mile Island accident
Power Output of Nuclear Reactor
Reaction rate R = (Neutron Flux)×(Cross-section)×(Number of Nuclei)
Flux φ = Neutrons m −2 s −1 Number of Nuclei = N
Cross-section σ = effective area, unit is barn = 10 −28 m 2
Example: Reactor core contains 10 4 kg of uranium enriched to 2%
in 235 U. Cross-section for neutron induced fission of 235 U = 579 barns.
Flux φ = 10 18 m −2 s −1 . Calculate the power output.
Number of 235 U nuclei = 10 4 (1000/238)(6 ×10 23 )(0.02) = 5.0×10 26
R = φσN = 10 18 ×579×10 −28 ×5.0×10 26 = 2.9×10 19 s −1
Energy per fission = 200 MeV = 200×10 6 ×1.6×10 −19 = 3.2×10 −11 J
So power output = 3.2×10 −11 ×2.9×10 19 = 0.93 GW th .
10 4 kg U(2%) 5.0×10 26 × 3.2 ×10 −11 J = 1.6 × 10 16 J ≡ 0.5 GW th y
Fast Breeder Reactors (FBR)
Predicted fossil reserves
~ 8.10 22 J
Fission reactors (thermal neutron) ~ 4.10 21 J
Fast breeder reactors (fast neutrons) ~ 2.10 23 J
fast breeder reactors are possible long-term solution to world’s
energy needs (~10 3 years) - ~50 times fission reactor energy reserve
Fission reactors consume 235 U so < 1% uranium utilised
Fast breeder reactors have small core of highly enriched fissile fuel
with no moderator. Emitted fast neutrons convert surrounding 238 U
to fissile 239 Pu quicker than fuel consumed by fast neutron induced
fission in core.
Basic reactions
β − β − α
n + 238 U 239 U 239 Np 239 Pu 235 U
t 1/2
=23.5m t 1/2
=2.35d t 1/2
=24,000y
‘fertile’ ie
spawns Pu
‘fissile’ ie
chain reaction

A Schematic of an ADSR
Use three
accelerators
for reliability
Energy Amplifier or Accelerator Driven Subcritical Reactor
World NUCLEAR POWER REACTORS 2003-04
Belgium
Canada*
China**
France
Germany
India
Japan
Korea (South)
Russia
Sweden
United Kingdom
USA
WORLD
billion
kWh
44.6
70.3
79.0
420.7
157.4
16.4
230.8
123.3
138.4
65.5
85.3
763.7
2525
% e
55
13
**
78
28
3
25
40
17
50
24
20
16
Operating
7
17
15
59
18
14
54
19
30
11
23
103
438
Building
0
1
4
0
0
9
3
1
5
0
0
1
28
Planned/
Proposed
0
2
26
0
0
24
12
8
9
0
0
0
106
Relative costs of electricity in the US (2003)
Costs of Electricity Generation (2003)
(25-year capital recovery, 85% lifetime capacity factor)
Source
Cents/kWe-hr
Nuclear 7.0
Coal 4.4
Gas 4.1
Nuclear Costs with reduced
Construction costs by 25% 5.8
Construction time by 12 months 5.6
Cost of capital ≡ coal and gas 4.7
With Carbon Tax $50/tC $100/tC $200/tC
Coal 5.6 6.8 9.2
Gas 4.6 5.1 6.2

Environmental Impact of Nuclear Power
Nuclear Fuel Cycle for typical reactor
Uranium
mining,
milling and
concentration
Final
disposaldeep
geological
depository)
4200tU as
enriched U 3 O 8
24t HLW
100m 3
600tU
in used
fuel
900m 3 in
containers
Conversion
to UF 6
(gas)
Reprocessing
and vitrification
of HLW
Interim storage
option (20 yrs +)
4200tU as
enriched UF 6
1
600tU
in used
fuel
2
Enrichment
to 3.5%
235
U
Reactor
Operation
1000 MW
30 years
operation
10,000 m 3 waste
(operating and
decommissioning)
600tU as
enriched
UF 6
Fuel
fabrication
as UO 2
600tU as fresh UO 2 fuel
200 10 9 kWh of
electricity
equivalent to
17.10 6 tonnes of oil
Categories of Nuclear Waste
1. LLW(Low Level Waste) ~89% of total volume
Low radioactivity, negligible long-lived activity (rags, tools, filters,
etc, from hospitals, research labs and nuclear power stations
2. ILW (Intermediate Level Waste) ~11% of total volume
Requires shielding, contains some long-lived activity (resins, sludges,
Fuel cladding) can be set in concrete/bitumen
3. HLW (High Level Waste) ~0.3% of total volume
Highly active, heat generating, long-lived activity requires vitrification
and long-term storage
National Waste Disposal Programmes
France: 400,000 m 3 of short-lived waste in shallow land burial at
La Manche site
Investigating sites for deep disposal of long-lived waste (including
vitrified HLW) from 2015
Germany :LLW and ILW in former salt mine
Investigations of Gorleben salt dome for final disposal of vitrified HLW
Japan: LLW put in shallow burial site (200,000m 3 capacity). HLW being
vitrified and stored for 30-50 years until suitable deep repository found
UK: Underground repository for LLW/ILW at Sellafield. HHW vitrified
stored 50 years at Sellafield before eventual disposal in deep repository
USA: Three LLW sites. National HLW site Yucca Mountain (?) (Nevada)
Outstanding Issues
• Deep repositories required to keep HLW intact for 10,000 years
Geological stability and water ingress are uncertain
• Long-term stability of vitrified waste unknown
• Public unease- easy target for anti-nuclear lobby
• Moral issue- should we burden future generations with our waste?
Counter argument: they will also need to dispose of nuclear waste, so
we are solving the technical problems for them. Also danger from
not reducing CO 2
Nuclear
Power
Chernobyl

Summary Nuclear
• Advantages: Low Carbon; Constant output;
Relatively cheap ~1.5x Fossil Fuels.
• Disadvantages: Perceived risk is high and concern
over radioactive waste disposal and proliferation
• Resource: 16 Mt U ~200 yrs at current output
(~300GW) [38 Mt U- including phosphates]
• Potential by 2050: ~ 1 TW e
• Increased resource using Th to breed 233 U
232
Th + n → 233 Th → 233 Pa (27d)→ 233 U
fertile
fissile
Resource ~ 2.5 Mt Th cf ~0.1 Mt 235 U
• Power density: ~ 1000 MW e per sq km