Nuclear Power Reactors
(Updated January 2015)
- Most nuclear electricity is generated using just two kinds of reactors which were developed in the 1950s and improved since.
- New designs are coming forward and some are in operation as
the first generation reactors come to the end of their operating lives.
- Around 13% of the world's electricity is produced from nuclear energy, more than from all sources worldwide in 1960.
This paper is about the main conventional types of nuclear reactor. For more advanced types, see
Advanced Reactors and
Small Reactors papers, and also
Generation IV reactors.
A nuclear reactor produces and controls the release of energy from
splitting the atoms of certain elements. In a nuclear power reactor, the
energy released is used as heat to make steam to generate electricity.
(In a research reactor the main purpose is to utilise the actual
neutrons produced in the core. In most naval reactors, steam drives a
turbine directly for propulsion.)
The principles for using nuclear power to produce electricity are the
same for most types of reactor. The energy released from continuous
fission of the atoms of the fuel is harnessed as heat in either a gas or
water, and is used to produce steam. The steam is used to drive the
turbines which produce electricity (as in most fossil fuel plants).
The world's first nuclear reactors operated naturally in a uranium
deposit about two billion years ago. These were in rich uranium
orebodies and moderated by percolating rainwater. The 17 known at Oklo
in west Africa, each less than 100 kW thermal, together consumed about
six tonnes of that uranium. It is assumed that these were not unique
worldwide.
Today, reactors derived from designs originally developed for
propelling submarines and large naval ships generate about 85% of the
world's nuclear electricity. The main design is the pressurised water
reactor (PWR) which has water at over 300°C under pressure in its
primary cooling/heat transfer circuit, and generates steam in a
secondary circuit. The less numerous boiling water reactor (BWR) makes
steam in the primary circuit above the reactor core, at similar
temperatures and pressure. Both types use water as both coolant and
moderator, to slow neutrons. Since water normally boils at 100°C, they
have robust steel pressure vessels or tubes to enable the higher
operating temperature. (Another type uses heavy water, with deuterium
atoms, as moderator. Hence the term ‘light water’ is used to
differentiate.)
Components of a nuclear reactor
There are several components common to most types of reactors:
Fuel. Uranium is the basic fuel. Usually pellets of uranium oxide (UO
2) are arranged in tubes to form fuel rods. The rods are arranged into fuel assemblies in the reactor core.*
*
In a new reactor with new fuel a neutron source is needed to get the
reaction going. Usually this is beryllium mixed with polonium, radium or
other alpha-emitter. Alpha particles from the decay cause a release of
neutrons from the beryllium as it turns to carbon-12. Restarting a
reactor with some used fuel may not require this, as there may be enough
neutrons to achieve criticality when control rods are removed.
Moderator. Material in the core which slows down the
neutrons released from fission so that they cause more fission. It is
usually water, but may be heavy water or graphite.
Control rods. These are made with neutron-absorbing
material such as cadmium, hafnium or boron, and are inserted or
withdrawn from the core to control the rate of reaction, or to halt
it.* In some PWR reactors, special control rods are used to enable the
core to sustain a low level of power efficiently. (Secondary control
systems involve other neutron absorbers, usually boron in the coolant –
its concentration can be adjusted over time as the fuel burns up.)
*
In
fission, most of the neutrons are released promptly, but some are
delayed. These are crucial in enabling a chain reacting system (or
reactor) to be controllable and to be able to be held precisely
critical.
Coolant. A fluid circulating through the core so as
to transfer the heat from it. In light water reactors the water
moderator functions also as primary coolant. Except in BWRs, there is
secondary coolant circuit where the water becomes steam. (See also later
section on primary coolant characteristics)
Pressure vessel or pressure tubes. Usually a robust
steel vessel containing the reactor core and moderator/coolant, but it
may be a series of tubes holding the fuel and conveying the coolant
through the surrounding moderator.
Steam generator. Part of the cooling system of
pressurised water reactors (PWR & PHWR) where the high-pressure
primary coolant bringing heat from the reactor is used to make steam for
the turbine, in a secondary circuit. Essentially a heat exchanger like a
motor car radiator*. Reactors have up to six 'loops', each with a steam
generator. Since 1980 over 110 PWR reactors have had their steam
generators replaced after 20-30 years service, 57 of these in USA.
* These are large heat exchangers for
transferring heat from one fluid to another – here from high-pressure
primary circuit in PWR to secondary circuit where water turns to steam.
Each structure weighs up to 800 tonnes and contains from 300 to 16,000
tubes about 2 cm diameter for the primary coolant, which is radioactive
due to nitrogen-16 (N-16, formed by neutron bombardment of oxygen, with
half-life of 7 seconds). The secondary water must flow through the
support structures for the tubes. The whole thing needs to be designed
so that the tubes don't vibrate and fret, operated so that deposits do
not build up to impede the flow, and maintained chemically to avoid
corrosion. Tubes which fail and leak are plugged, and surplus capacity
is designed to allow for this. Leaks can be detected by monitoring N-16
levels in the steam as it leaves the steam generator.
Containment. The structure around the reactor and
associated steam generators which is designed to protect it from outside
intrusion and to protect those outside from the effects of radiation in
case of any serious malfunction inside. It is typically a metre-thick
concrete and steel structure.
There are several different types of reactors as indicated in the following Table.
Nuclear power plants in commercial operation
| Reactor type |
Main Countries |
Number |
GWe |
Fuel |
Coolant |
Moderator |
| Pressurised Water Reactor (PWR) |
US, France, Japan, Russia, China
|
273
|
253
|
enriched UO2
|
water
|
water
|
| Boiling Water Reactor (BWR) |
US, Japan, Sweden
|
81
|
76
|
enriched UO2
|
water
|
water
|
| Pressurised Heavy Water Reactor 'CANDU' (PHWR) |
Canada
|
48
|
24
|
natural UO2
|
heavy water
|
heavy water
|
| Gas-cooled Reactor (AGR & Magnox) |
UK
|
15
|
8
|
natural U (metal),
enriched UO2
|
CO2
|
graphite
|
| Light Water Graphite Reactor (RBMK & EGP) |
Russia
|
11 + 4
|
10.2
|
enriched UO2
|
water
|
graphite
|
| Fast Neutron Reactor (FBR) |
Russia
|
2
|
0.6
|
PuO2 and UO2
|
liquid sodium
|
none
|
| |
TOTAL |
434 |
372 |
|
|
|
IAEA data, end of 2013. GWe = capacity in thousands of megawatts (gross)
Source: Nuclear Engineering International Handbook 2011, updated to 1/1/12For reactors under construction: see paper
Plans for New Reactors Worldwide.
Fuelling a nuclear power reactor
Most reactors need to be shut down for refuelling, so that the
pressure vessel can be opened up. In this case refuelling is at
intervals of 1-2 years, when a quarter to a third of the fuel assemblies
are replaced with fresh ones. The CANDU and RBMK types have pressure
tubes (rather than a pressure vessel enclosing the reactor core) and can
be refuelled under load by disconnecting individual pressure tubes.
If graphite or heavy water is used as moderator, it is possible to
run a power reactor on natural instead of enriched uranium. Natural
uranium has the same elemental composition as when it was mined (0.7%
U-235, over 99.2% U-238), enriched uranium has had the proportion of the
fissile isotope (U-235) increased by a process called enrichment,
commonly to 3.5 - 5.0%. In this case the moderator can be ordinary
water, and such reactors are collectively called light water reactors.
Because the light water absorbs neutrons as well as slowing them, it is
less efficient as a moderator than heavy water or graphite.
During operation, some of the U-238 is changed to
plutonium, and Pu-239 ends up providing about one third of the energy from the fuel.
In most reactors the fuel is ceramic uranium oxide (UO
2
with a melting point of 2800°C) and most is enriched. The fuel pellets
(usually about 1 cm diameter and 1.5 cm long) are typically arranged in a
long zirconium alloy (zircaloy) tube to form a fuel rod, the zirconium
being hard, corrosion-resistant and permeable to neutrons.* Numerous
rods form a fuel assembly, which is an open lattice and can be lifted
into and out of the reactor core. In the most common reactors these are
about 3.5 to 4 metres long.
*Zirconium is an important mineral for nuclear power,
where it finds its main use. It is therefore subject to controls on
trading. It is normally contaminated with hafnium, a neutron absorber,
so very pure 'nuclear grade' Zr is used to make the zircaloy, which is
about 98% Zr plus about 1.5% tin, also iron, chromium and sometimes
nickel to enhance its strength.
Burnable poisons are often used (especially in BWR) in fuel or
coolant to even out the performance of the reactor over time from fresh
fuel being loaded to refuelling. These are neutron absorbers which decay
under neutron exposure, compensating for the progressive build up of
neutron absorbers in the fuel as it is burned. The best known is
gadolinium, which is a vital ingredient of fuel in naval reactors where
installing fresh fuel is very inconvenient, so reactors are designed to
run more than a decade between refuellings.
The power rating of a nuclear power reactor
Nuclear power plant reactor power outputs are quoted in three ways:
- Thermal MWt, which depends on the design of the actual nuclear
reactor itself, and relates to the quantity and quality of the steam it
produces.
- Gross electrical MWe indicates the power produced by the attached
steam turbine and generator, and also takes into account the ambient
temperature for the condenser circuit (cooler means more electric power,
warmer means less). Rated gross power assumes certain conditions with
both.
- Net electrical MWe, which is the power available to be sent out from
the plant to the grid, after deducting the electrical power needed to
run the reactor (cooling and feed-water pumps, etc.) and the rest of the
plant.*
*
Net electrical MWe and gross MWe vary
slightly from summer to winter, so normally the lower summer figure, or
an average figure, is used. If the summer figure is quoted plants may
show a capacity factor greater than 100% in cooler times. Watts Bar PWR
in Tennessee is reported to run at about 1125 MWe in summer and about
1165 MWe net in winter, due to different condenser cooling water
temperatures. Some design options, such as powering the main large
feed-water pumps with electric motors (as in EPR) rather than steam
turbines (taking steam before it gets to the main turbine-generator),
explains some gross to net differences between different reactor types.
The EPR has a relatively large drop from gross to net MWe for this
reason.
The relationship between these is expressed in two ways:
- Thermal efficiency %, the ratio of gross MWe to thermal MW. This
relates to the difference in temperature between the steam from the
reactor and the cooling water. It is often 33-37%.
- Net efficiency %, the ratio of net MWe achieved to thermal MW. This is a little lower, and allows for plant usage.
In WNA papers and figures and WNN items, generally net MWe is used
for operating plants, and gross MWe for those under construction or
planned/proposed.
Pressurised Water Reactor (PWR)
This is the most common type, with over 230 in use for power
generation and several hundred more employed for naval propulsion. The
design of PWRs originated as a
submarine power plant.
PWRs use ordinary water as both coolant and moderator. The design is
distinguished by having a primary cooling circuit which flows through
the core of the reactor under very high pressure, and a secondary
circuit in which steam is generated to drive the turbine. In Russia
these are known as VVER types – water-moderated and -cooled.
A PWR has fuel assemblies of 200-300 rods each, arranged vertically
in the core, and a large reactor would have about 150-250 fuel
assemblies with 80-100 tonnes of uranium.
Water in the reactor core reaches about 325°C, hence it must be kept
under about 150 times atmospheric pressure to prevent it boiling.
Pressure is maintained by steam in a pressuriser (see diagram). In the
primary cooling circuit the water is also the moderator, and if any of
it turned to steam the fission reaction would slow down. This negative
feedback effect is one of the safety features of the type. The secondary
shutdown system involves adding boron to the primary circuit.
The secondary circuit is under less pressure and the water here boils
in the heat exchangers which are thus steam generators. The steam
drives the turbine to produce electricity, and is then condensed and
returned to the heat exchangers in contact with the primary circuit.
Boiling Water Reactor (BWR)
This design has many similarities to the PWR, except that there is
only a single circuit in which the water is at lower pressure (about 75
times atmospheric pressure) so that it boils in the core at about 285°C.
The reactor is designed to operate with 12-15% of the water in the top
part of the core as steam, and hence with less moderating effect and
thus efficiency there. BWR units can operate in load-following mode
more readily then PWRs.
The steam passes through drier plates (steam separators) above the
core and then directly to the turbines, which are thus part of the
reactor circuit. Since the water around the core of a reactor is always
contaminated with traces of radionuclides, it means that the turbine
must be shielded and radiological protection provided during
maintenance. The cost of this tends to balance the savings due to the
simpler design. Most of the radioactivity in the water is very
short-lived*, so the turbine hall can be entered soon after the reactor
is shut down.
* mostly N-16, with a 7 second half-life
A BWR fuel assembly comprises 90-100 fuel rods, and there are up to
750 assemblies in a reactor core, holding up to 140 tonnes of uranium.
The secondary control system involves restricting water flow through the
core so that more steam in the top part reduces moderation.
Pressurised Heavy Water Reactor (PHWR or CANDU)
The PHWR reactor design has been developed since the 1950s in Canada
as the CANDU, and more recently also in India. PHWRs generally use
natural uranium (0.7% U-235) oxide as fuel, hence needs a more efficient
moderator, in this case heavy water (D
2O).** The PHWR
produces more energy per kilogram of mined uranium than other
designs, but also produces a much larger amount of used fuel per unit
output.
** with the CANDU system, the moderator is enriched (i.e. water) rather than the fuel – a cost trade-off.
The moderator is in a large tank called a calandria, penetrated by
several hundred horizontal pressure tubes which form channels for the
fuel, cooled by a flow of heavy water under high pressure in the primary
cooling circuit, reaching 290°C. As in the PWR, the primary coolant
generates steam in a secondary circuit to drive the turbines. The
pressure tube design means that the reactor can be refuelled
progressively without shutting down, by isolating individual pressure
tubes from the cooling circuit. It is also less costly to build than
designs with a large pressure vessel, but the tubes have not proved as
durable.
A CANDU fuel assembly consists of a bundle of 37 half metre long fuel
rods (ceramic fuel pellets in zircaloy tubes) plus a support structure,
with 12 bundles lying end to end in a fuel channel. Control rods
penetrate the calandria vertically, and a secondary shutdown system
involves adding gadolinium to the moderator. The heavy water moderator
circulating through the body of the calandria vessel also yields some
heat (though this circuit is not shown on the diagram above).
Newer PHWR designs such as the Advanced Candu Reactor (ACR) have light water cooling and slightly-enriched fuel.
CANDU reactors can accept a variety of fuels. They may be run on
recycled uranium from reprocessing LWR used fuel, or a blend of this and
depleted uranium left over from enrichment plants. About 4000 MWe of
PWR might then fuel 1000 MWe of CANDU capacity, with addition of
depleted uranium. Thorium may also be used in fuel.
Advanced Gas-cooled Reactor (AGR)
These are the second generation of British gas-cooled reactors, using
graphite moderator and carbon dioxide as primary coolant. The fuel is
uranium oxide pellets, enriched to 2.5-3.5%, in stainless steel tubes.
The carbon dioxide circulates through the core, reaching 650°C and then
past steam generator tubes outside it, but still inside the concrete and
steel pressure vessel (hence 'integral' design). Control rods penetrate
the moderator and a secondary shutdown system involves injecting
nitrogen to the coolant.
The AGR was developed from the
Magnox reactor, also graphite moderated and CO
2
cooled, and one of these is still operating in UK to late 2014. They
use natural uranium fuel in metal form. Secondary coolant is water.
Light water graphite-moderated reactor (RBMK)
This is a Soviet design, developed from plutonium production
reactors. It employs long (7 metre) vertical pressure tubes running
through graphite moderator, and is cooled by water, which is allowed to
boil in the core at 290°C, much as in a BWR. Fuel is low-enriched
uranium oxide made up into fuel assemblies 3.5 metres long. With
moderation largely due to the fixed graphite, excess boiling simply
reduces the cooling and neutron absorbtion without inhibiting the
fission reaction, and a positive feedback problem can arise, which is
why they have never been built outside the Soviet Union. See appendix on
RBMK Reactors for more detail.
Advanced reactors
Several generations of reactors are commonly distinguished.
Generation I reactors were developed in 1950-60s and only one is still
running today. They mostly used natural uranium fuel and used graphite
as moderator. Generation II reactors are typified by the present US
fleet and most in operation elsewhere. They typically use enriched
uranium fuel and are mostly cooled and moderated by water. Generation
III are the Advanced Reactors evolved from these, the first few of which
are in operation in Japan and others are under construction and ready
to be ordered. They are developments of the second generation with
enhanced safety. There is no clear distinction Gen II to Gen III.
Generation IV designs are still on the drawing board and will not be
operational before 2020 at the earliest, probably later. They will tend
to have closed fuel cycles and burn the long-lived actinides now forming
part of spent fuel, so that fission products are the only high-level
waste. Of seven designs under development, 4 or 5 will be fast neutron
reactors. Four will use fluoride or liquid metal coolants, hence operate
at low pressure. Two will be gas-cooled. Most will run at much higher
temperatures than today’s water-cooled reactors. See
Generation IV Reactors paper.
More than a dozen (Generation III)
advanced reactor
designs are in various stages of development. Some are evolutionary
from the PWR, BWR and CANDU designs above, some are more radical
departures. The former include the Advanced Boiling Water Reactor, a few
of which are now operating with others under construction. The
best-known radical new design has the fuel as large 'pebbles' and uses
helium as coolant, at very high temperature, possibly to drive a turbine
directly.
Considering the closed fuel cycle, Generation 1-3 reactors recycle
plutonium (and possibly uranium), while Generation IV are expected to
have full actinide recycle.
Fast neutron reactors (FNR)
Some reactors (only one in commercial service) do not have a
moderator and utilise fast neutrons, generating power from plutonium
while making more of it from the U-238 isotope in or around the fuel.
While they get more than 60 times as much energy from the original
uranium compared with the normal reactors, they are expensive to build.
Further development of them is likely in the next decade, and the main
designs expected to be built in two decades are FNRs. If they are
configured to produce more fissile material (plutonium) than they
consume they are called Fast Breeder Reactors (FBR). See also
Fast Neutron Reactors and
Small Reactors papers.
Floating nuclear power plants
Apart from over 200 nuclear reactors powering various kinds of ships,
Rosatom in Russia has set up a subsidiary to supply floating nuclear
power plants ranging in size from 70 to 600 MWe. These will be mounted
in pairs on a large barge, which will be permanently moored where it is
needed to supply power and possibly some desalination to a shore
settlement or industrial complex. The first has two 40 MWe reactors
based on those in icebreakers and will operate at a remote site in
Siberia. Electricity cost is expected to be much lower than from present
alternatives.
The Russian KLT-40S is a reactor well proven in icebreakers and now
proposed for wider use in desalination and, on barges, for remote area
power supply. Here a 150 MWt unit produces 35 MWe (gross) as well as up
to 35 MW of heat for desalination or district heating. These are
designed to run 3-4 years between refuelling and it is envisaged that
they will be operated in pairs to allow for outages, with on-board
refuelling capability and used fuel storage. At the end of a 12-year
operating cycle the whole plant is taken to a central facility for
2-year overhaul and removal of used fuel, before being returned to
service. Two units will be mounted on a 21,000 tonne barge. A larger
Russian factory-built and barge-mounted reactor is the VBER-150, of 350
MW thermal, 110 MWe. The larger VBER-300 PWR is a 325 MWe unit,
originally envisaged in pairs as a floating nuclear power plant,
displacing 49,000 tonnes. As a cogeneration plant it is rated at 200 MWe
and 1900 GJ/hr. See also
Nuclear Power in Russia paper.
Lifetime of nuclear reactors
Most of today's nuclear plants which were originally designed for 30
or 40-year operating lives. However, with major investments in systems,
structures and components lives can be extended, and in several
countries there are active programs to extend operating lives. In the
USA most of the more than one hundred reactors are expected to be
granted licence extensions from 40 to 60 years. This justifies
significant capital expenditure in upgrading systems and components,
including building in extra performance margins.
Some components simply wear out, corrode or degrade to a low level of
efficiency. These need to be replaced. Steam generators are the most
prominent and expensive of these, and many have been replaced after
about 30 years where the reactor otherwise has the prospect of running
for 60 years. This is essentially an economic decision. Lesser
components are more straightforward to replace as they age. In Candu
reactors, pressure tube replacement has been undertaken on some plants
after about 30 years operation.
A second issue is that of obsolescence. For instance, older reactors
have analogue instrument and control systems. Thirdly, the properties of
materials may degrade with age, particularly with heat and neutron
irradiation. In respect to all these aspects, investment is needed to
maintain reliability and safety. Also, periodic safety reviews are
undertaken on older plants in line with international safety conventions
and principles to ensure that safety margins are maintained.
Another important issue is knowledge management (KM) over the full
lifecycle from design, through construction and operation to
decommissioning for reactors and other facilities. This may span a
century and involve several countries, and involve a succession of
companies. The plant lifespan will cover several generations of
engineers. Data needs to be transferable across several generations of
software and IT hardware, as well as being shared with other operators
of similar plants.* Significant modifications may be made to the design
over the life of the plant, so original documentation is not sufficient,
and loss of design base knowledge can have huge implications (eg
Pickering A and Bruce A in Ontario). Knowledge management is often a
shared responsibility and is essential for effective decision-making and
the achievement of plant safety and economics.
See also section on Ageing, in
Safety of Nuclear Power Reactors paper.
Load-following capacity
Nuclear power plants are essentially base-load generators, running
continuously. This is because their power output cannot readily be
ramped up and down on a daily and weekly basis, and in this respect they
are similar to most coal-fired plants. (It is also uneconomic to run
them at less than full capacity, since they are expensive to build but
cheap to run.) However, in some situations it is necessary to vary the
output according to daily and weekly load cycles on a regular basis, for
instance in France, where there is a very high reliance on nuclear
power.
While BWRs can be made to follow loads reasonably easily without
burning the core unevenly, this is not as readily achieved in a PWR. The
ability of a PWR to run at less than full power for much of the time
depends on whether it is in the early part of its 18 to 24-month
refueling cycle or late in it, and whether it is designed with special
control rods which diminish power levels throughout the core without
shutting it down. Thus, though the ability on any individual PWR reactor
to run on a sustained basis at low power decreases markedly as it
progresses through the refueling cycle, there is considerable scope for
running a fleet of reactors in load-following mode. See further
information in the
Nuclear Power in France paper.
As fast neutron reactors become established in future years, their ability to load-follow will be a benefit.
Primary coolants
The advent of some of the designs mentioned above provides
opportunity to review the various primary coolants used in nuclear
reactors. There is a wide variety – gas, water, light metal, heavy metal
and salt:
Water or heavy water must be maintained at very high
pressure (1000-2200 psi, 7-15 MPa, 150 atmospheres) to enable it to
function well above 100°C, up to 345°C, as in present reactors. This has
a major influence on reactor engineering. However, supercritical water
around 25 MPa can give 45% thermal efficiency – as at some fossil-fuel
power plants today with outlet temperatures of 600°C, and at ultra
supercritical levels (30+ MPa) 50% may be attained.
Water cooling of steam condensers is fairly standard in all power
plants, because it works very well, it is relatively inexpensive, and
there is a huge experience base. Water is a lot more effective than air
for removing heat.
Helium must be used at similar pressure (1000-2000
psi, 7-14 MPa) to maintain sufficient density for efficient operation.
Again, there are engineering implications, but it can be used in the
Brayton cycle to drive a turbine directly.
Carbon dioxide was used in early British reactors,
and their current AGRs which operate at much higher temperatures than
light water reactors. It is denser than helium and thus likely to give
better thermal conversion efficiency. It also leaks less readily than
helium. There is now interest in supercritical CO
2 for the Brayton cycle.
Sodium, as normally used in fast neutron reactors at
around 550ºC, melts at 98°C and boils at 883°C at atmospheric pressure,
so despite the need to keep it dry the engineering required to contain
it is relatively modest. It has high thermal conductivity. However,
normally water/steam is used in the secondary circuit to drive a turbine
(Rankine cycle) at lower thermal efficiency than the Brayton cycle. In
some designs sodium is in a secondary circuit to steam generators.
Sodium does not corrode the metals used in the fuel cladding or primary
circuit, nor the fuel itself if there is cladding damage, but it is very
reactive generally. In particular it reacts exothermically with water
or steam to liberate hydrogen. Sodium has a low neutron capture cross
section, but it is enough for some Na-23 to become Na-24, which is a
beta-emitter and gamma-active with 15-year half-life, so some shielding
is required. If a reactor needs to be shut down frequently, NaK eutectic
which is liquid at room temperature (about 13°C) may be used as
coolant, but the potassium is pyrophoric, which increases the hazard.
Lead or lead-bismuth eutectic in fast neutron
reactors are capable of higher temperature operation. They are
transparent to neutrons, aiding efficiency due to greater spacing
between fuel pins which then allows coolant flow by convection for decay
heat removal, and since they do not react with water the heat exchanger
interface is safer. They do not burn when exposed to air. However, they
are corrosive of fuel cladding and steels, which originally limited
temperatures to 550°C. With today's materials 650°C can be reached, and
in future 800°C is envisaged with the second stage of Gen IV
development, using oxide dispersion-strengthened steels. They have much
higher thermal conductivity than water, but lower than sodium. A problem
is that Pb-Bi yields toxic polonium (Po-210) activation products. Pb-Bi
melts at a relatively low 125°C (hence eutectic) and boils at 1670°C,
Pb melts at 327°C and boils at 1737°C but is very much more abundant and
cheaper to produce than bismuth, hence is envisaged for large-scale use
in the future, though freezing must be prevented. The development of
nuclear power based on Pb-Bi cooled fast neutron reactors is likely to
be limited to a total of 50-100 GWe, basically for small reactors in
remote places. In 1998 Russia declassified a lot of research information
derived from its experience with submarine reactors, and US interest in
using Pb or Pb-Bi for small reactors has increased subsequently. The
Gen4 Module (Hyperion) reactor will use lead-bismuth eutectic which is
45% Pb, 55% Bi. A secondary circuit generating steam is likely.
Fluoride salts boil at around 1400°C at atmospheric
pressure, so allow several options for use of the heat, including using
helium in a secondary Brayton cycle circuit with thermal efficiencies of
48% at 750°C to 59% at 1000°C, or manufacture of hydrogen. Fluoride
salts have very low vapour pressure even at red heat, carry more heat
than the same volume of water, have good heat transfer properties, have
low neutron absorbtion, good neutron moderation capability, are not
damaged by radiation, are chemically very stable so absorb all fission
products well and do not react violently with air or water, and some are
inert to some common structural metals. Some gamma-active F-20 is
formed by neutron capture, but has very short half-life (11 seconds).
Lithium-beryllium fluoride Li
2BeF
4 (FLiBe) salt
is a eutectic version of LiF (2LiF + BeF2) which solidifies at 459°C
and boils at 1430°C. It is favoured in MSR and AHTR/FHR primary cooling
and when uncontaminated has a low corrosion effect. LiF without the
toxic beryllium solidifies at about 500°C and boils at about 1200°C.
FLiNaK (LiF-NaF-KF) is also eutectic and solidifies at 454°C and boils
at 1570°C. It has a higher neutron cross-section than FLiBe or LiF but
can be used intermediate cooling loops.
All low-pressure liquid coolants allow all their
heat to be delivered at high temperatures, since the temperature drop in
heat exchangers is less than with gas coolants. Also, with a good
margin between operating and boiling temperatures, passive cooling for
decay heat is readily achieved. Since heat exchangers do leak to some
small extent, having incompatible primary and secondary coolants can be a
problem. The less pressure difference across the heat exchanger, the
less is the problem.
The removal of
passive decay heat is a vital feature
of primary cooling systems, beyond heat transfer to do work. When the
fission process stops, fission product decay continues and a substantial
amount of heat is added to the core. At the moment of shutdown, this is
about 6.5% of the full power level, but after an hour it drops to about
1.5% as the short-lived fission products decay. After a day, the decay
heat falls to 0.4%, and after a week it will be only 0.2%. This heat
could melt the core of a light water reactor unless it is reliably
dissipated, as shown in 2011 at Fukushima, where about 1.5% of the heat
was being generated when the tsunami disabled the cooling. In passive
systems, some kind of convection flow is relied upon.
The top AHTR/FHR line is potential, lower one practical today. See also paper on
Cooling Power Plants.
There is some radioactivity in the cooling water flowing through the
core of a water-cooled reactor, due mainly to the activation product
nitrogen-16, formed by neutron capture from oxygen. N-16 has a half-life
on only 7 seconds but produces high-energy gamma radiation during
decay. It is the reason that access to a BWR turbine hall is restricted
during actual operation.
Nuclear reactors for process heat
Producing steam to drive a turbine and generator is relatively easy,
and a light water reactor running at 350°C does this readily. As the
above section and Figure show, other types of reactor are required for
higher temperatures. A 2010 US Department of Energy document quotes
500°C for a liquid metal cooled reactor (FNR), 860°C for a molten salt
reactor (MSR), and 950°C for a high temperature gas-cooled reactor
(HTR). Lower-temperature reactors can be used with supplemental gas
heating to reach higher temperatures, though employing an LWR would not
be practical or economic. The DOE said that high reactor outlet
temperatures in the range 750 to 950°C were required to satisfy all end
user requirements evaluated to date for the Next Generation Nuclear
Plant.
Primitive reactors
The world's oldest known nuclear reactors operated at what is now
Oklo in Gabon, West Africa. About 2 billion years ago, at least 17
natural nuclear reactors achieved criticality in a rich deposit of
uranium ore. Each operated intermittently at about 20 kW thermal, the
reaction ceasing whenever the water turned to steam so that it ceased to
function as moderator. At that time the concentration of U-235 in all
natural uranium was about three percent instead of 0.7 percent as at
present. (U-235 decays much faster than U-238, whose half-life is about
the same as the age of the Earth.) These natural chain reactions,
started spontaneously with the presence of water acting as a moderator,
continued overall for about 2 million years before finally dying
away. It appears that each reactor operated in pulses of about 30
minutes – interrupted when the water turned to steam, thereby switching
it off for a few hours until it cooled. It is estimated that about 130
TWh of heat was produced. (The reactors were discovered when assays of
mined uranium showed only 0.717% U-235 instead of 0.720% as everywhere
else on the planet. Further investigation identified significant
concentrations of fission products from both uranium and plutonium.)
During this long reaction period about 5.4 tonnes of fission products
as well as up to two tonnes of plutonium together with other
transuranic elements were generated in the orebody. The initial
radioactive products have long since decayed into stable elements but
close study of the amount and location of these has shown that there was
little movement of radioactive wastes during and after the nuclear
reactions. Plutonium and the other transuranics remained immobile.
Sources: Wilson, P.D., 1996, The Nuclear Fuel Cycle, OUP.
Scientific American 2005
article on Oklo
Source: Nuclear Regulatory Commission
