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. Those at Oklo in west Africa, each less than 100 kW thermal, together consumed about six tonnes of that uranium.
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.(not
in BWR) Part of the cooling system 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 may have up to four 'loops', each with a
steam generator.
* 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
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 tin, 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.
he 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. 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 kg of mined uranium than other designs.
** with the CANDU system, the moderator is enriched (ie 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.
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 readily 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 can 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) to enable
it to function above 100°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 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. 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.
Lead or lead-bismuth eutectic
in fast neutron reactors are capable of higher temperature operation.
They are transparent to neutrons, aiding efficiency, 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 700°C is in sight, using
oxide dispersion-strengthened steels. 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.
Molten fluoride salt
boils at 1400°C at atmospheric pressure, so allows 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, have reasonably good heat transfer properties, are not
damaged by radiation, do not react violently with air or water, and are
inert to some common structural metals.
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.
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.
Top AHTR 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 3.7 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 by the presence of water acting as a moderator, continued
overall for about 2 million years before finally dying away.
During this long reaction period about 5.4 tonnes of fission products
as well as 1.5 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.
