NUCLEAR ENERGY:
Nuclear Energy is the
use of sustained Nuclear fission to generate heat and do useful
work. Nuclear Electric Plants, Nuclear Ships and Submarines use controlled
nuclear energy to heat water and produce steam, while in
space, nuclear energy decays naturally in a radioisotope thermoelectric
generator. Scientists are experimenting with fusion
energy for future generation, but these experiments do not currently generate
useful energy.
Nuclear power provides about 6% of
the world's energy and 13–14% of the world's electricity, with the U.S., France, and Japan together accounting for about 50% of
nuclear generated electricity. Also, more than 150 naval vessels using nuclear propulsion have been built.
Nuclear power is controversial and
there is an ongoing debate about the use of nuclear energy. Proponents, such as
the World Nuclear Association and IAEA, contend that nuclear power
is a sustainable energy source that reduces carbon
emissions. Opponents, such as Greenpeace International and NIRS, believe that nuclear
power poses many threats to people and the environment.
Some
serious nuclear and radiation
accidents have occurred. Nuclear power plant accidents include the Chernobyl disaster (1986), Fukushima I nuclear accidents (2011),
and the Three Mile Island accident (1979).[10]
Nuclear-powered submarine mishaps include
the K-19 reactor accident (1961), the K-27 reactor accident (1968), and
the K-431 reactor accident (1985). International
research is continuing into safety improvements such as passively
safe plants, and the possible future use of nuclear
fusion.
NUCLEAR
FISSION:
In nuclear
physics and nuclear chemistry, nuclear fission is a nuclear
reaction in which the nucleus of an atom splits into smaller parts
(lighter nuclei), often producing free neutrons and photons (in the
form of gamma
rays). The two nuclei produced are most often of comparable size, typically
with a mass ratio around 3:2 for common fissile isotopes.[1][2]
Most fissions are binary fissions, but occasionally (2 to 4 times per 1000
events), three positively-charged fragments are produced in a ternary fission.
The smallest of these ranges in size from a proton to an argon nucleus.
Fission is usually an
energetic nuclear reaction induced by a neutron, although it
is occasionally seen as a form of spontaneous radioactive
decay, especially in very high-mass-number isotopes. The unpredictable
composition of the products (which vary in a broad probabilistic and somewhat
chaotic manner) distinguishes fission from purely quantum-tunnelling processes
such as proton emission, alpha decay
and cluster
decay, which give the same products every time.
Fission of heavy elements
is an exothermic reaction which can release large
amounts of energy
both as electromagnetic radiation and as kinetic
energy of the fragments (heating the bulk material where fission takes place). In order
for fission to produce energy, the total binding energy of the resulting elements must be
less than that of the starting element. Fission is a form of nuclear transmutation because the resulting
fragments are not the same element
as the original atom.
NUCLEAR FUSION:
In nuclear
physics, nuclear chemistry and astrophysics
nuclear fusion is the process by which two or more atomic
nuclei join together, or "fuse", to form a single heavier
nucleus. This is usually accompanied by the release or absorption of large
quantities of energy.
Large-scale thermonuclear fusion processes, involving many nuclei fusing at
once, must occur in matter at very high densities and temperatures.
The fusion of two nuclei with
lower masses than iron
(which, along with nickel,
has the largest binding energy per nucleon) generally releases
energy while the fusion of nuclei heavier than iron absorbs energy. The
opposite is true for the reverse process, nuclear
fission.
In the simplest case of hydrogen
fusion, two protons must be brought close enough for the weak
nuclear force to convert either of the identical protons into a neutron,
thus forming the hydrogen isotope deuterium. In
more complex cases of heavy ion fusion involving two or more nucleons, the reaction
mechanism is different, but the same result occurs— smaller nuclei are
combined into larger nuclei.
Nuclear fusion occurs naturally in
all active stars.
Synthetic fusion as a result of human actions has also been achieved, although
this has not yet been completely controlled as a source of nuclear
power (see: fusion power). In the laboratory,
successful nuclear physics experiments have been carried out that involve the
fusion of many different varieties of nuclei, but the energy output has been
negligible in these studies. In fact, the amount of energy put into the process
has always exceeded the energy output.
Uncontrolled nuclear fusion has been
carried out many times in nuclear weapons testing, which results in a
deliberate explosion.
These explosions have always used the heavy isotopes of hydrogen,
deuterium (H-2) and tritium (H-3), and never the much more common isotope of
hydrogen (H-1), sometimes called "protium".
Building upon the nuclear transmutation experiments by Ernest
Rutherford, carried out several years earlier, the fusion of the light
nuclei (hydrogen isotopes) was first accomplished by Mark
Oliphant in 1932. Then, the steps of the main cycle of nuclear fusion in
stars were first worked out by Hans Bethe throughout the remainder of that decade.
Research into fusion for military
purposes began in the early 1940s as part of the Manhattan
Project, but this was not accomplished until 1951 (see the Greenhouse
Item nuclear test), and nuclear fusion on a large scale in an explosion was
first carried out on November 1, 1952, in the Ivy Mike hydrogen
bomb test. Research into developing controlled thermonuclear fusion for
civil purposes also began in the 1950s, and it continues to this day.
TYPES OF
REACTORS:
Pressurized
water reactors (PWRs) constitute a majority of all western nuclear power plants and are one of two types
of light water reactor (LWR), the other type being
boiling water reactors (BWRs). In a PWR the primary coolant (water) is pumped
under high pressure to the reactor core where it is heated by the energy
generated by the fission of atoms. The heated water then flows to a
steam generator where it transfers its thermal energy to a secondary system
where steam is generated and flows to turbines which, in turn, spins an
electric generator. In contrast to a boiling water reactor, pressure in the
primary coolant loop prevents the water from boiling within the reactor. All
LWRs use ordinary light
water as both coolant and neutron
moderator.
PWRs
were originally designed to serve as nuclear propulsion for nuclear
submarines and were used in the original design of the second commercial
power plant at Shippingport Atomic Power Station.
PWRs
currently operating in the United States are considered Generation II reactors. Russia's VVER reactors are
similar to U.S. PWRs. France operates many PWRs to generate the
bulk of their electricity
Several
hundred PWRs are used for marine propulsion in aircraft
carriers, nuclear submarines and ice
breakers. In the US, they were originally designed at the Oak Ridge National Laboratory for use
as a nuclear submarine power plant. Follow-on work was conducted by
Westinghouse Bettis Atomic Power Laboratory.[1]
The first commercial nuclear power plant at Shippingport Atomic Power Station
was originally designed as a pressurized water reactor, on insistence from Admiral Hyman
G. Rickover that a viable commercial plant would include none of the
"crazy thermodynamic cycles that everyone else wants to build."
The US Army Nuclear Power Program operated
pressurized water reactors from 1954 to 1974.
Three Mile Island Nuclear
Generating Station initially operated two pressurized water reactor plants,
TMI-1 and TMI-2. The partial meltdown of TMI-2 in 1979
essentially ended the growth in new construction nuclear power plants in the
United States.
Design
Pictorial explanation of power transfer in a pressurized water reactor.
Primary coolant is in orange and the secondary coolant (steam and later
feedwater) is in blue.
Nuclear
fuel in the reactor vessel is engaged in a fission chain reaction, which produces heat,
heating the water in the primary coolant loop by thermal conduction through the
fuel cladding. The hot primary coolant is pumped into a heat
exchanger called the steam generator, where it flows
through hundreds or thousands of tubes (usually 3/4 inch in diameter).
Heat is transferred through the walls of these tubes to the lower pressure
secondary coolant located on the sheet side of the exchanger where it
evaporates to pressurized steam. The transfer of heat is accomplished without
mixing the two fluids, which is desirable since the primary coolant might
become radioactive. Some common steam generator arrangements are u-tubes or
single pass heat exchangers. In a nuclear power station, the pressurized steam
is fed through a steam turbine which drives an electrical generator connected to the electric
grid for distribution. After passing through the turbine the secondary coolant
(water-steam mixture) is cooled down and condensed in a condenser. The condenser converts the
steam to a liquid so that it can be pumped back into the steam generator, and
maintains a vacuum at the turbine outlet so that the pressure drop across the
turbine, and hence the energy extracted from the steam, is maximized. Before
being fed into the steam generator, the condensed steam (referred to as
feedwater) is sometimes preheated in order to minimize thermal shock.
The steam generated has other uses besides
power generation. In nuclear ships and submarines, the steam is fed through a
steam turbine connected to a set of speed reduction gears to a shaft used
for propulsion. Direct mechanical action by
expansion of the steam can be used for a steam-powered aircraft
catapult or similar applications. District
heating by the steam is used in some countries and direct heating is
applied to internal plant applications.
Two
things are characteristic for the pressurized water reactor (PWR) when compared with other reactor types:
coolant loop separation from the steam system and pressure inside the primary
coolant loop. In a PWR, there are
two separate coolant loops (primary and secondary), which are both filled with
demineralized/deionized water. A boiling water reactor, by contrast, has only
one coolant loop, while more exotic designs such as breeder
reactors use substances other than water for coolant and moderator (e.g.
sodium in its liquid state as coolant or graphite as a moderator). The pressure
in the primary coolant loop is typically 15–16 megapascals
(150–160 bar),
which is notably higher than in other nuclear
reactors, and nearly twice that of a boiling water reactor (BWR). As an
effect of this, only localized boiling occurs and steam will recondense
promptly in the bulk fluid. By contrast, in a boiling water reactor the primary
coolant is designed to boil.
PWR Reactor Design
PWR Reactor
Vessel
Coolant
Light water is used
as the primary coolant in a PWR.
It enters the bottom of the reactor core at about 275 °C
(530 °F)
and is heated as it flows upwards through the reactor core to a temperature of
about 315 °C (600 °F). The
water remains liquid despite the high temperature due to the high pressure in
the primary coolant loop, usually around 155 bar
(15.5 MPa
153 atm, 2,250 psig). In water, the critical point occurs at around
647 K
(374 °C or
705 °F) and
22.064 MPa (3200 PSIA or 218 atm).[7]
Pressure
in the primary circuit is maintained by a pressurizer,
a separate vessel that is connected to the primary circuit and partially filled
with water which is heated to the saturation temperature (boiling point) for
the desired pressure by submerged electrical heaters. To achieve a pressure of
155 bar, the pressurizer temperature is maintained at 345 °C,
which gives a subcooling margin (the difference between the pressurizer
temperature and the highest temperature in the reactor core) of 30 °C.
Thermal transients in the reactor coolant system result in large swings in
pressurizer liquid volume, total pressurizer volume is designed around
absorbing these transients without uncovering the heaters or emptying the
pressurizer. Pressure transients in the primary coolant system manifest as
temperature transients in the pressurizer and are controlled through the use of
automatic heaters and water spray, which raise and lower pressurizer
temperature, respectively.
To achieve maximum heat transfer, the primary
circuit temperature, pressure and flow rate are arranged such that subcooled nucleate
boiling takes place as the coolant passes over the nuclear fuel rods.
The
coolant is pumped around the primary circuit by powerful pumps, which can
consume up to 6 MW each. After picking up heat as it passes through the reactor
core, the primary coolant transfers heat in a steam generator to water in a
lower pressure secondary circuit, evaporating the secondary coolant to
saturated steam — in most designs 6.2 MPa (60 atm, 900 psia), 275 °C
(530 °F) — for use in the steam turbine. The cooled primary coolant is
then returned to the reactor vessel to be heated again.
Moderator
Pressurized
water reactors, like all thermal reactor designs, require the fast fission
neutrons to be slowed down (a process called moderation or thermalization) in
order to interact with the nuclear fuel and sustain the chain reaction. In PWRs
the coolant water is used as a moderator
by letting the neutrons undergo multiple collisions with light hydrogen atoms
in the water, losing speed in the process. This "moderating" of
neutrons will happen more often when the water is denser (more collisions will
occur). The use of water as a moderator is an important safety feature of PWRs,
as an increase in temperature may cause the water to turn to steam - thereby
reducing the extent to which neutrons are slowed down and hence reducing the
reactivity in the reactor. Therefore, if reactivity increases beyond normal,
the reduced moderation of neutrons will cause the chain reaction to slow down,
producing less heat. This property, known as the negative temperature coefficient of reactivity,
makes PWR reactors very stable.
In
contrast, the RBMK
reactor design used at Chernobyl, which uses graphite instead of water as the
moderator and uses boiling water as the coolant, has a large positive thermal
coefficient of reactivity, that increases heat generation when coolant water
temperatures increase. This makes the RBMK design less stable than pressurized
water reactors. In addition to its property of slowing down neutrons when
serving as a moderator, water also has a property of absorbing neutrons, albeit
to a lesser degree. When the coolant water temperature increases, the boiling
increases, which creates voids. Thus there is less water to absorb thermal
neutrons that have already been slowed down by the graphite moderator, causing
an increase in reactivity. This property is called the void
coefficient of reactivity, and in an RBMK reactor like Chernobyl, the void
coefficient is positive, and fairly large, causing rapid transients. This
design characteristic of the RBMK reactor is generally seen as one of several
causes of the Chernobyl accident.[10]
Heavy water
has very low neutron absorption, so heavy water reactors such as CANDU
reactors also have a positive void coefficient, though it is not as large
as that of an RBMK like Chernobyl; these reactors are designed with a number of
safety systems not found in the original RBMK design, which are designed to
handle or react to this as needed.
PWRs are
designed to be maintained in an undermoderated state, meaning that there is
room for increased water volume or density to further increase moderation,
because if moderation were near saturation, then a reduction in density of the
moderator/coolant could reduce neutron absorption significantly while reducing
moderation only slightly, making the void coefficient positive. Also, light
water is actually a somewhat stronger moderator of neutrons than heavy water,
though heavy water's neutron absorption is much lower. Because of these two
facts, light water reactors have a relatively small moderator volume and
therefore have compact cores. One next generation design, the supercritical water reactor, is even
less moderated. A less moderated neutron energy spectrum does worsen the
capture/fission ratio for 235U and especially 239Pu,
meaning that more fissile nuclei fail to fission on neutron absorption and
instead capture the neutron to become a heavier nonfissile isotope, wasting one
or more neutrons and increasing accumulation of heavy transuranic actinides,
some of which have long half-lives.
Fuel
PWR fuel
bundle This fuel bundle is from a pressurized water reactor of
the nuclear passenger and cargo ship NS Savannah.
Designed and built by the Babcock and Wilcox Company.
After
enrichment the uranium dioxide (UO2) powder is fired in a
high-temperature, sintering furnace to create hard, ceramic pellets of
enriched uranium dioxide. The cylindrical pellets are then clad in a
corrosion-resistant zirconium metal alloy Zircaloy which
are backfilled with helium to aid heat conduction and detect leakages. Zircaloy is
chosen because of its mechanical properties and its low absorption cross
section. The finished fuel rods are grouped in fuel assemblies, called fuel
bundles, that are then used to build the core of the reactor. A typical PWR has
fuel assemblies of 200 to 300 rods each, and a large reactor would have about
150–250 such assemblies with 80–100 tonnes of uranium in all. Generally, the
fuel bundles consist of fuel rods bundled 14 × 14 to 17 × 17.
A PWR produces on the order of 900 to 1,500 MWe. PWR fuel
bundles are about 4 meters in length.
Refuelings for most commercial PWRs is on an
18–24 month cycle. Approximately one third of the core is replaced each
refueling, though some more modern refueling schemes may reduce refuel time to
a few days and allow refueling to occur on a shorter periodicity.
Control
In PWRs
reactor power can be viewed as following steam (turbine) demand due to the
reactivity feedback of the temperature change caused by increased or decreased
steam flow. (See: Negative temperature coefficient.)
Boron and control rods are used to maintain primary system temperature at the
desired point. In order to decrease power, the operator throttles shut turbine
inlet valves. This would result in less steam being drawn from the steam
generators. This results in the primary loop increasing in temperature. The
higher temperature causes the reactor to fission less and decrease in power.
The operator could then add boric acid and/or insert control rods to decrease
temperature to the desired point.
Reactivity
adjustment to maintain 100% power as the fuel is burned up in most commercial
PWRs is normally achieved by varying the concentration of boric acid
dissolved in the primary reactor coolant. Boron readily absorbs neutrons and
increasing or decreasing its concentration in the reactor coolant will
therefore affect the neutron activity correspondingly. An entire control system
involving high pressure pumps (usually called the charging and letdown system)
is required to remove water from the high pressure primary loop and re-inject
the water back in with differing concentrations of boric acid. The reactor
control rods, inserted through the reactor vessel head directly into the fuel
bundles, are moved for the following reasons:
·
To start up the reactor.
·
To shut down the primary nuclear
reactions in the reactor.
·
To accommodate short term transients
such as changes to load on the turbine.
The
control rods can also be used:
·
To compensate for nuclear poison
inventory.
·
To compensate for nuclear
fuel depletion.
but
these effects are more usually accommodated by altering the primary coolant
boric acid concentration.
In
contrast, BWRs have no
boron in the reactor coolant and control the reactor power by adjusting the
reactor coolant flow rate.
Advantages:
PWR
reactors are very stable due to their tendency to produce less power as
temperatures increase; this makes the reactor easier to operate from a
stability standpoint as long as the post shutdown period of 1 to 3 years[citation needed] has pumped
cooling.
PWR turbine
cycle loop is separate from the primary loop, so the water in the secondary
loop is not contaminated by radioactive materials.
PWRs can
passively scram the reactor in the event that offsite power is lost to immediately
stop the primary nuclear reaction. The control rods are held by electromagnets
and fall by gravity when current is lost; full insertion safely shuts down the
primary nuclear reaction. However, nuclear reactions of the fission products
continue to generate decay heat at initially roughly 7% of full power level,
which requires 1 to 3 years of water pumped cooling. If cooling fails during
this post-shutdown period, the reactor can still overheat and meltdown. Upon
loss of coolant the decay heat can raise the rods above 2200 degrees Celsius,
where upon the hot Zirconium alloy metal used for casing the nuclear
fuel rods spontaneously explodes in contact with the cooling water or steam,
which leads to the separation of water in to its constituent elements (hydrogen and oxygen). In this
event there is a high danger of hydrogen explosions, threatening structural
damage and/or the exposure of highly radioactive stored fuel rods in the
vicinity outside the plant in pools (approximately 15 tons of fuel is
replenished each year to maintain normal PWR
operation).
Disadvantages
The coolant
water must be highly pressurized to remain liquid at high temperatures. This
requires high strength piping and a heavy pressure vessel and hence increases
construction costs. The higher pressure can increase the consequences of a loss-of-coolant accident.[14]
The reactor pressure vessel is manufactured from ductile steel but, as the
plant is operated, neutron flux from the reactor causes this steel to become
less ductile. Eventually the ductility of the steel will reach limits determined by the
applicable boiler and pressure vessel standards, and the pressure vessel must
be repaired or replaced. This might not be practical or economic, and so
determines the life of the plant.
Additional high
pressure components such as reactor coolant pumps, pressurizer, steam
generators, etc. are also needed. This also increases the capital cost and
complexity of a PWR power plant.
The high
temperature water coolant with boric acid dissolved in it is corrosive to carbon
steel (but not stainless steel); this can cause radioactive
corrosion products to circulate in the primary coolant loop. This not only
limits the lifetime of the reactor, but the systems that filter out the corrosion
products and adjust the boric acid concentration add significantly to the
overall cost of the reactor and to radiation exposure. Occasionally, this has
resulted in severe corrosion to control rod drive mechanisms when the boric
acid solution leaked through the seal between the mechanism itself and the
primary system.
Natural uranium
is only 0.7% uranium-235, the isotope necessary for thermal reactors. This
makes it necessary to enrich the uranium fuel, which increases the costs of
fuel production. If heavy water is used, it is possible to operate the
reactor with natural uranium, but the production of heavy water requires large
amounts of energy and is hence expensive.
Because water
acts as a neutron moderator, it is not possible to build a fast neutron reactor with a PWR design. A reduced moderation water reactor
may however achieve a breeding ratio greater than unity, though this
reactor design has disadvantages of its own.
Boiling
Water Reactor:
The boiling
water reactor (BWR) is a type of light water nuclear
reactor used for the generation of electrical power. It is the second most
common type of electricity-generating nuclear reactor after the pressurized water reactor (PWR), also a
type of light water nuclear reactor. The BWR was developed by the Idaho National Laboratory and General
Electric in the mid-1950s. The main present manufacturer is GE Hitachi Nuclear Energy, which
specializes in the design and construction of this type of reactor.
Early concepts
The BWR
concept was developed slightly later than the PWR
concept. Development of the BWR started in the early 1950s, and was a
collaboration between GE and several US national laboratories.
Research
into nuclear power in the US was led by the 3 military services. The Navy,
seeing the possibility of turning submarines into full-time underwater
vehicles, and ships that could steam around the world without refueling, sent
their man in engineering, Captain Hyman
Rickover to run their nuclear power program. Rickover decided on the PWR
route for the Navy, as the early researchers in the field of nuclear power
feared that the direct production of steam within a reactor would cause
instability, while they knew that the use of pressurized water would
definitively work as a means of heat transfer. This concern led to the US's
first research effort in nuclear power being devoted to the PWR, which was
highly suited for naval vessels (submarines, especially), as space was at a
premium, and PWRs could be made compact and high-power enough to fit in such,
in any event.
But
other researchers wanted to investigate whether the supposed instability caused
by boiling water in a reactor core would really cause instability. In particular,
Samuel Untermyer II, a researcher at Idaho National Laboratory (INL), proposed
and oversaw a series of experiments: the BORAX
experiments—to see if a boiling water reactor would be feasible for use in
energy production. He found that it was, after subjecting his reactors to quite
strenuous tests, proving the safety principles of the BWR.
Following
this series of tests, GE got involved and collaborated with INL to bring this technology to market.
Larger-scale tests were conducted through the late 1950s/early/mid-1960s that
only partially used directly-generated (primary) nuclear boiler system steam to
feed the turbine and incorporated heat exchangers for the generation of
secondary steam to drive separate parts of the turbines. The literature does
not indicate why this was the case, but it was eliminated on production models
of the BWR.
First series of production BWRs (BWR/1–BWR/6)
The
first generation of production boiling water reactors saw the incremental
development of the unique and distinctive features of the BWR: the torus (used to quench
steam in the event of a transient requiring the quenching of steam), as well as
the drywell, the elimination of the heat exchanger, the steam dryer, the
distinctive general layout of the reactor building, and the standardization of
reactor control and safety systems. The first, General Electric, series of
production BWRs evolved through 6 iterative design phases, each termed BWR/1
through BWR/6. (BWR/4s, BWR/5s, and BWR/6s are the most common types in service
today.) The vast majority of BWRs in service throughout the world belong to one
of these design phases.
1st generation
BWR: BWR/1 with Mark I containment.
2nd generation
BWRs: BWR/2, BWR/3 and some BWR/4 with Mark I containment. Other BWR/4, and BWR/5
with Mark-II containment.
3rd generation
BWRs: BWR/6 with Mark-III containment.
Browns Ferry Unit 1 drywell and
wetwell under construction, a BWR/4 using the Mark I containment
Containment
variants were constructed using either concrete or steel for the Primary
Containment, Drywell and Wetwell in various combinations.[5]
Apart
from the GE designs there were others by ABB, MITSU, Toshiba and KWU. See List of boiling water reactors.
The advanced boiling water reactor (ABWR)
A newer
design of BWR is known as the Advanced Boiling Water Reactor
(ABWR). The ABWR was developed in the late 1980s and early 1990s, and has been
further improved to the present day. The ABWR incorporates advanced
technologies in the design, including computer control, plant automation,
control rod removal, motion, and insertion, in-core pumping, and nuclear safety
to deliver improvements over the original series of production BWRs, with a
high power output (1350 MWe per reactor), and a significantly lowered
probability of core damage. Most significantly, the ABWR was a completely
standardized design, that could be made for series production.[citation needed]
The ABWR
was approved by the U.S. Nuclear Regulatory Commission for production as a
standardized design in the early 1990s. Subsequently, numerous ABWRs were built
in Japan. One development spurred by the success of the ABWR in Japan is that
GE's nuclear energy division merged with Hitachi Corporation's nuclear energy
division, forming GE Hitachi, who is now the major worldwide developer of the
BWR design.
The simplified boiling water reactor (SBWR)
General
Electric (GE) also developed a different concept for a new boiling water
reactor (BWR) at the same time as the ABWR, known as the simplified boiling
water reactor (SBWR). This smaller (600 megawatt electrical (MWe) per reactor) was notable for its
incorporation—for the first time ever in a light water reactor—of "passive
safety" design principles. The concept of passive safety means that the
reactor, rather than requiring the intervention of active systems, such as
emergency injection pumps, to keep the reactor within safety margins, was
instead designed to return to a safe state solely through operation of natural
forces if a safety-related contingency developed.
For
example, if the reactor got too hot, it would trigger a system that would release
soluble neutron absorbers (generally a solution of borated materials, or a
solution of borax),
or materials that greatly hamper a chain reaction by absorbing neutrons, into
the reactor core. The tank containing the soluble neutron absorbers would be
located above the reactor, and the absorption solution, once the system was
triggered, would flow into the core through force of gravity, and bring the
reaction to a near-complete stop. Another example was the Isolation Condenser
system, which relied on the principle of hot water/steam rising to bring hot
coolant into large heat exchangers located above the reactor in very deep tanks
of water, thus accomplishing residual heat removal. Yet another example was the
omission of recirculation pumps within the core; these pumps were used in other
BWR designs to keep cooling water moving; they were expensive, hard to reach to
repair, and could occasionally fail; so as to improve reliability, the ABWR incorporated
no less than 10 of these recirculation pumps, so that even if several failed, a
sufficient number would remain serviceable so that an unscheduled shutdown
would not be necessary, and the pumps could be repaired during the next
refueling outage. Instead, the designers of the Simplified Boiling Water
Reactor used thermal analysis to design the reactor core such that natural
circulation (cold water falls, hot water rises) would bring water to the center
of the core to be boiled.
The
ultimate result of the passive safety features of the SBWR would be a reactor
that would not require human intervention in the event of a major safety
contingency for at least 48 hours following the safety contingency; thence, it
would only require periodic refilling of cooling water tanks located completely
outside of the reactor, isolated from the cooling system, and designed to
remove reactor waste heat through evaporation. The Simplified Boiling Water
Reactor was submitted to the United States Nuclear Regulatory Commission,
however, it was withdrawn prior to approval; still, the concept remained
intriguing to General Electric's designers, and served as the basis of future
developments.
The economic simplified boiling water reactor (ESBWR)
During a
period beginning in the late 1990s, GE engineers proposed to combine the
features of the advanced boiling water reactor design with the distinctive
safety features of the simplified boiling water reactor design, along with
scaling up the resulting design to a larger size of 1,600 MWe (4,500 MWth).
This Economic Simplified Boiling
Water Reactor design has been submitted to the U.S. Nuclear Regulatory
Commission for approval, and the subsequent Final Design Review is near
completion.
Reportedly,
this design has been advertised as having a core damage probability of only 3×10−8
core damage events per reactor-year.[citation needed] (That is,
there would need to be 3 million ESBWRs operating before one would expect a
single core-damaging event during their 100-year lifetimes. Earlier designs of
the BWR (the BWR/4) had core damage probabilities as high as 1×10−5
core-damage events per reactor-year.)[6]
This extraordinarily low CDP for the ESBWR far exceeds the other large LWRs on
the market.
Advantages and disadvantages
Advantages
·
The reactor vessel and associated
components operate at a substantially lower pressure (about 75 times
atmospheric pressure) compared to a PWR
(about 158 times atmospheric pressure).
·
Pressure vessel is subject to
significantly less irradiation compared to a PWR,
and so does not become as brittle with age.
·
Operates at a lower nuclear fuel
temperature.
·
Fewer components due to no steam
generators and no pressurizer vessel. (Older BWRs have external recirculation
loops, but even this piping is eliminated in modern BWRs, such as the ABWR.)
·
Lower risk (probability) of a rupture
causing loss of coolant compared to a PWR,
and lower risk of core damage should such a rupture occur. This is due to fewer
pipes, fewer large diameter pipes, fewer welds and no steam generator tubes.
·
NRC assessments of limiting fault
potentials indicate if such a fault occurred, the average BWR would be less
likely to sustain core damage than the average PWR
due to the robustness and redundancy of the Emergency Core Cooling System (ECCS).
·
Unlike PWRs, BWRs have at least a few
steam-turbine driven ECCS systems
that can be directly operated by steam produced after a reactor shutdown, and
require no electrical power. This results in less dependence on emergency
diesel generators—of which there are four—in any event.
·
Measuring the water level in the
pressure vessel is the same for both normal and emergency operations, which
results in easy and intuitive assessment of emergency conditions.
·
Can operate at lower core power density
levels using natural circulation without forced flow.
·
A BWR may be designed to operate using
only natural circulation so that recirculation pumps are eliminated entirely.
(The new ESBWR design uses natural circulation.)
·
BWRs do not use boric acid
to control fission burn-up, leading to less possibility of corrosion within the
reactor vessel and piping. (Corrosion from boric acid must be carefully
monitored in PWRs; it has been demonstrated that reactor vessel head corrosion
can occur if the reactor vessel head is not properly maintained. See Davis-Besse.
Since BWRs do not utilize boric acid, these contingencies are eliminated.)
·
BWRs generally have N-2
redundancy on their major safety-related systems, which normally consist of
four "trains" of components. This generally means that up to two of
the four components of a safety system can fail and the system will still
perform if called upon.
·
Due to their single major vendor
(GE/Hitachi), the current fleet of BWRs have predictable, uniform designs that,
while not completely standardized, generally are very similar to one another.
The ABWR/ESBWR designs are completely standardized. Lack of standardization
remains a problem with PWRs, as, at least in the United States, there are three
design families represented among the current PWR fleet (Combustion
Engineering, Westinghouse, and Babcock & Wilcox), within these families,
there are quite divergent designs.
·
Additional families of PWRs are being
introduced. For example, Mitsubishi's APWR,
Areva's US-EPR, and Westinghouse's AP1000/AP600 will add
diversity and complexity to an already diverse crowd, and possibly cause
customers seeking stability and predictability to seek other designs, such as
the BWR.
·
BWRs are overrepresented in imports, if
the importing nation doesn't have a nuclear navy (PWRs are favored by nuclear
naval states due to their compact, high-power design used on nuclear-powered
vessels; since naval reactors are generally not exported, they cause national
skill to be developed in PWR
design, construction, and operation), or special national aspirations (special
national aspirations lead to a marked preference for the CANDU reactor type due to
special features of that type). This may be due to the fact that BWRs are
ideally suited for peaceful uses like power generation,
process/industrial/district heating, and desalinization, due to low cost,
simplicity, and safety focus, which come at the expense of larger size and
slightly lower thermal efficiency.
·
Sweden is standardized mainly on BWRs.
·
Mexico's only two reactors are BWRs.
·
Japan experimented with both PWRs and
BWRs, but most builds as of late have been of BWRs, specifically ABWRs.
·
In the CEGB open competition in the
early 1960s for a standard design for UK 2nd-generation power reactors, the PWR didn't even make it to the final round, which
was a showdown between the BWR (preferred for its easily understood design as
well as for being predictable and "boring") and the AGCR, a uniquely British design; the
indigenous design won, possibly on technical merits, possibly due to the
proximity of a general election.
Disadvantages
·
Much larger pressure vessel than for a PWR of similar power, with correspondingly higher
cost. (However, the overall cost is reduced because a modern Complex
calculations for managing consumption of nuclear fuel during operation due to
"two phase (water and steam) fluid flow" in the upper part of the
core. This requires more instrumentation in the reactor core. The innovation of
computers, however, makes this less of an issue.
·
BWR has no main steam generators and
associated piping.)
·
Contamination of the turbine by
short-lived activation products. This means that shielding
and access control around the steam turbine are required during normal
operations due to the radiation levels arising from the steam entering directly
from the reactor core. This is a moderately minor concern, as most of the
radiation flux is due to Nitrogen-16, which has a half-life measured in
seconds, allowing the turbine chamber to be entered into within minutes of
shutdown.
·
Though the present fleet of BWRs are
said to be less likely to suffer core damage from the "1 in 100,000
reactor-year" limiting fault than the present fleet of PWRs are (due to
increased ECCS robustness and
redundancy) there have been concerns raised about the pressure containment
ability of the as-built, unmodified Mark I containment – that such may be
insufficient to contain pressures generated by a limiting fault combined with
complete ECCS failure that results in extremely severe core damage. In this
double failure scenario, assumed to be extremely unlikely prior to the Fukushima I nuclear accidents, an
unmodified Mark I containment can allow some degree of radioactive release to
occur. This is supposed to be mitigated by the modification of the Mark I
containment; namely, the addition of an outgas stack system that, if
containment pressure exceeds critical setpoints, is supposed to allow the
orderly discharge of pressurizing gases after the gases pass through activated
carbon filters designed to trap radionuclides.[7]
·
A BWR requires active cooling for a
period of several hours to several days following shutdown, depending on its
power history. Full insertion of BWRs control rods safely shuts down the
primary nuclear reaction. However, radioactive decay of the fission products in
the fuel will continue to actively generate decay heat
at a gradually decreasing rate, requiring pumping of cooling water for an
initial period to prevent overheating of the fuel. If active cooling fails
during this post-shutdown period, the reactor can still overheat to a
temperature high enough that zirconium in the fuel cladding will react with
water and steam, producing hydrogen gas. In this event there is a high danger
of hydrogen explosions, threatening structural damage to the reactor and/or
associated safety systems and/or the exposure of highly radioactive spent fuel
rods that may be stored in the reactor building (approx 15 tons of fuel is
replenished each year to maintain normal BWR operation) as happened with the Fukushima I nuclear accidents.
·
Control rods are inserted from below for
current BWR designs. There are two available hydraulic power sources that can
drive the control rods into the core for a BWR under emergency conditions.
There is a dedicated high pressure hydraulic accumulator and also the pressure
inside of the reactor pressure vessel available to each control rod. Either the
dedicated accumulator (one per rod) or reactor pressure is capable of fully
inserting each rod. Most other reactor types use top entry control rods that
are held up in the withdrawn position by electromagnets, causing them to fall
into the reactor by gravity if power is lost.
NUCLEAR
WASTE AND ITS DISPOSAL:
Radioactive
waste is a waste product containing radioactive
material. It is usually the product of a nuclear process such as nuclear
fission, though industries not directly connected to the nuclear
power industry may also produce radioactive waste.
Radioactivity
diminishes over time,
so in principle the waste needs to be isolated for a period of time until it no
longer poses a hazard.
This can mean hours to years for some common medical or industrial radioactive
wastes, or thousands of years for high-level wastes from nuclear power plants and nuclear
weapons reprocessing.
The
majority of radioactive waste is "low-level
waste", meaning it has low levels of radioactivity per mass or volume.
The main
approaches to managing radioactive waste to date have been segregation and
storage for short-lived wastes, near-surface disposal for low and some intermediate
level wastes, and deep burial or transmutation for the long-lived, high-level
wastes.
A
summary of the amounts of radioactive wastes and management approaches for most
developed countries are presented and reviewed periodically as part of the IAEA Joint Convention
on Safety of Spent Fuel Management and the Safety of Radioactive Waste
Management.
Types of radioactive waste
Although
not significantly radioactive, uranium mill tailings are waste. They are
byproduct material from the rough processing of uranium-bearing ore. They are
sometimes referred to as 11(e)2 wastes, from the section of the U.S. Atomic
Energy Act that defines them. Uranium mill tailings typically also contain
chemically hazardous heavy metals such as lead and arsenic. Vast
mounds of uranium mill tailings are left at many old mining sites, especially
in Colorado,
New Mexico,
and Utah.
Low
level waste (LLW) is generated from hospitals and
industry, as well as the nuclear fuel cycle. It comprises paper, rags, tools,
clothing, filters, etc., which contain small amounts of mostly short-lived
radioactivity. Commonly, LLW is designated as such as a precautionary measure
if it originated from any region of an 'Active Area', which frequently includes
offices with only a remote possibility of being contaminated with radioactive
materials. Such LLW typically exhibits no higher radioactivity than one would
expect from the same material disposed of in a non-active area, such as a
normal office block. Some high activity LLW requires shielding during handling
and transport but most LLW is suitable for shallow land burial. To reduce its
volume, it is often compacted or incinerated before disposal. Low level waste
is divided into four classes, class A, B, C and GTCC, which means "Greater
Than Class C".
Intermediate
level waste (ILW) contains higher amounts of radioactivity and in
some cases requires shielding. ILW includes resins, chemical sludge and metal
reactor fuel
cladding, as well as contaminated materials from reactor decommissioning. It
may be solidified in concrete or bitumen for disposal. As a general rule,
short-lived waste (mainly non-fuel materials from reactors) is buried in
shallow repositories, while long-lived waste (from fuel and fuel-reprocessing)
is deposited in deep underground facilities. U.S. regulations
do not define this category of waste; the term is used in Europe and elsewhere.
Spent Fuel Flasks are transported by railway in the United Kingdom. Each
flask is constructed of 14 in (360 mm) thick solid steel and weighs
in excess of 50 tons
High
level waste (HLW) is produced by nuclear
reactors. It contains fission
products and transuranic elements generated in the reactor
core. It is highly radioactive and often thermally hot. HLW accounts for
over 95% of the total radioactivity produced in the process of nuclear electricity generation. The amount of HLW
worldwide is currently increasing by about 12,000 metric tons every year, which
is the equivalent to about 100 double-decker buses or a two-story structure
with a footprint the size of a basketball court. A 1000-MWe nuclear power plant
produces about 27 tonnes of spent nuclear fuel (unreprocessed) every year.
Transuranic
waste (TRUW) as defined by U.S. regulations is, without regard
to form or origin, waste that is contaminated with alpha-emitting transuranic
radionuclides with half-lives greater than 20 years, and concentrations greater
than 100 nCi/g
(3.7 MBq/kg),
excluding High Level Waste. Elements that have an atomic
number greater than uranium are called transuranic ("beyond
uranium"). Because of their long half-lives, TRUW is disposed more
cautiously than either low level or intermediate level waste. In the US it
arises mainly from weapons production, and consists of clothing, tools, rags,
residues, debris and other items contaminated with small amounts of radioactive
elements (mainly plutonium).
Under US
law, transuranic waste is further categorized into "contact-handled"
(CH) and "remote-handled" (RH) on the basis of radiation dose
measured at the surface of the waste container. CH TRUW has a surface dose rate
not greater than 200 mrem per hour (2 mSv/h),
whereas RH TRUW has a surface dose rate of 200 mrem per hour (2 mSv/h) or greater. CH TRUW
does not have the very high radioactivity of high level waste, nor its high
heat generation, but RH TRUW can be highly radioactive, with surface dose rates
up to 1000000 mrem per hour (10000 mSv/h). The US
currently permanently disposes of defense-related TRUW at the Waste Isolation Pilot Plant.
Preventing of Waste
Due to
the many advances in reactor design, it is today possible to reduce the
radioactive waste by a factor 100. This can be done by using new reactor types
such as Generation_IV_reactor. This reducion of
nuclear waste is possible these new reactor types are capable of burning the
lower actinides.
Management of Waste
Modern medium to high level transport container for nuclear waste.
See also: High-level radioactive waste
management, List of nuclear waste
treatment technologies, and Environmental effects of nuclear
power
Of
particular concern in nuclear waste management are two long-lived fission
products, Tc-99 (half-life 220,000 years) and I-129 (half-life 17 million
years), which dominate spent fuel radioactivity after a few thousand years. The
most troublesome transuranic elements in spent fuel are Np-237 (half-life two
million years) and Pu-239 (half life 24,000 years). Nuclear waste requires
sophisticated treatment and management to successfully isolate it from
interacting with the biosphere. This usually necessitates treatment, followed by
a long-term management strategy involving storage, disposal or transformation
of the waste into a non-toxic form. Governments around the world are
considering a range of waste management and disposal options, though there has
been limited progress toward long-term waste management solutions.
Initial treatment of waste
Vitrification
Long-term
storage of radioactive waste requires the stabilization of the waste into a
form which will neither react nor degrade for extended periods of time. One way
to do this is through vitrification.[23]
Currently at Sellafield the high-level waste (PUREX first cycle raffinate) is
mixed with sugar
and then calcined. Calcination involves passing the waste through a heated,
rotating tube. The purposes of calcination are to evaporate the water from the
waste, and de-nitrate the fission products to assist the stability of the glass
produced.
The 'calcine' generated is fed continuously
into an induction heated furnace with fragmented glass.[25]
The resulting glass is a new substance in which the waste products are bonded
into the glass matrix when it solidifies. This product, as a melt, is poured
into stainless steel cylindrical containers
("cylinders") in a batch process. When cooled, the fluid solidifies
("vitrifies") into the glass. Such glass, after being formed, is
highly resistant to water.
After filling a cylinder, a seal is welded onto the
cylinder. The cylinder is then washed. After being inspected for external
contamination, the steel cylinder is stored, usually in an underground
repository. In this form, the waste products are expected to be immobilized for
a long period of time (many thousands of years).
The glass inside a cylinder is usually a black
glossy substance. All this work (in the United Kingdom) is done using hot cell systems.
The sugar is added to control the ruthenium
chemistry and to stop the formation of the volatile RuO4 containing radioactive ruthenium
isotopes. In the west, the glass is normally a borosilicate glass (similar to Pyrex), while in the
former Soviet
bloc it is normal to use a phosphate
glass. The amount of fission products in the glass must be limited because
some (palladium,
the other Pt group metals, and tellurium) tend to form metallic phases which separate from
the glass. Bulk vitrification uses electrodes to melt soil and wastes, which
are then buried underground.[28]
In Germany a vitrification plant is in use; this is treating the waste from a
small demonstration reprocessing plant which has since been closed down.[24][29]
Ion exchange
It is
common for medium active wastes in the nuclear industry to be treated with ion
exchange or other means to concentrate the radioactivity into a small
volume. The much less radioactive bulk (after treatment) is often then
discharged. For instance, it is possible to use a ferric hydroxide floc to
remove radioactive metals from aqueous mixtures.[30]
After the radioisotopes are absorbed onto the ferric hydroxide, the resulting
sludge can be placed in a metal drum before being mixed with cement to form a
solid waste form.[31]
In order to get better long-term performance (mechanical stability) from such
forms, they may be made from a mixture of fly ash, or blast
furnace slag,
and Portland cement, instead of normal concrete (made
with Portland cement, gravel and sand).
Synroc
The
Australian Synroc
(synthetic rock) is a more sophisticated way to immobilize such waste, and this
process may eventually come into commercial use for civil wastes (it is currently
being developed for US military wastes). Synroc was invented by the late Prof
Ted Ringwood (a geochemist) at the Australian National University.[32]
The Synroc contains pyrochlore and cryptomelane type minerals. The original
form of Synroc (Synroc C) was designed for the liquid high level waste (PUREX raffinate) from a light water reactor. The main minerals in this
Synroc are hollandite (BaAl2Ti6O16), zirconolite
(CaZrTi2O7) and perovskite
(CaTiO3). The zirconolite and perovskite are hosts for the actinides.
The strontium
and barium will
be fixed in the perovskite. The caesium will be fixed in the hollandite.
Long term management of Waste
The time
frame in question when dealing with radioactive waste ranges from 10,000 to
1,000,000 years, according to studies based on the effect of estimated
radiation doses. Researchers suggest that forecasts of health detriment for
such periods should be examined critically. Practical studies only consider up
to 100 years as far as effective planning and cost evaluations are concerned.
Long term behavior of radioactive wastes remains a subject for ongoing research
projects.
Geologic disposal
The
process of selecting appropriate deep final repositories for high level
waste and spent fuel is now under way in several countries (Schacht
Asse II and the Waste Isolation Pilot Plant) with the first expected to be
commissioned some time after 2010. The basic concept is to locate a large,
stable geologic formation and use mining technology to excavate a tunnel, or
large-bore tunnel boring machines (similar to those used
to drill the Channel Tunnel from England to France) to drill a
shaft 500–1,000 meters below the surface where rooms or vaults can be excavated
for disposal of high-level radioactive waste. The goal is to permanently
isolate nuclear waste from the human environment. Many people remain
uncomfortable with the immediate stewardship cessation of this disposal
system, suggesting perpetual management and monitoring would be more prudent.
Because
some radioactive species have half-lives longer than one million years, even
very low container leakage and radionuclide migration rates must be taken into
account. Moreover, it may require more than one half-life until some nuclear
materials lose enough radioactivity to cease being lethal to living things. A
1983 review of the Swedish radioactive waste disposal program by the National
Academy of Sciences found that country’s estimate of several hundred thousand
years—perhaps up to one million years—being necessary for waste isolation
“fully justified.” Storing high level nuclear waste above ground for a century
or so is considered appropriate by many scientists. This allows the material to
be more easily observed and any problems detected and managed, while decay of
radionuclides over this time period significantly reduces the level of
radioactivity and associated harmful effects to the container material. It is
also considered likely that over the next century newer materials will be
developed which will not break down as quickly when exposed to a high neutron
flux, thus increasing the longevity of the container once it is permanently
buried.
Sea-based
options for disposal of radioactive waste[42]
include burial beneath a stable abyssal
plain, burial in a subduction zone that would slowly carry the waste downward
into the Earth's mantle, and burial beneath a remote
natural or human-made island. While these approaches all have merit and would
facilitate an international solution to the problem of disposal of radioactive
waste, they would require an amendment of the Law of the Sea.
Article
1 (Definitions), 7., of the 1996 Protocol to the Convention on the Prevention
of Marine Pollution by Dumping of Wastes and Other Matter, (the London Dumping
Convention) states:
“Sea” means all marine waters other than the internal waters of States, as
well as the seabed and the subsoil thereof; it does not include sub-seabed
repositories accessed only from land.”
The
proposed land-based subductive waste disposal method disposes of nuclear waste
in a subduction
zone accessed from land,
and therefore is not prohibited by international agreement. This method has
been described as the most viable means of disposing of radioactive waste, and
as the state-of-the-art as of 2001 in nuclear waste disposal technology.
Another approach termed Remix & Return would blend high-level waste with uranium
mine and mill tailings down to the level of the original radioactivity of
the uranium
ore, then replace it in inactive uranium mines. This approach has the
merits of providing jobs for miners who would double as disposal staff, and of
facilitating a cradle-to-grave cycle for radioactive materials, but would be
inappropriate for spent reactor fuel in the absence of reprocessing, due to the
presence in it of highly toxic radioactive elements such as plutonium.
Deep borehole disposal is the concept of
disposing of high-level radioactive waste from nuclear reactors in extremely
deep boreholes. Deep borehole disposal seeks to place the waste as much as five
kilometers beneath the surface of the Earth and relies primarily on the immense
natural geological barrier to confine the waste safely and permanently so that
it should never pose a threat to the environment.
Transmutation
There
have been proposals for reactors that consume nuclear waste and transmute it to
other, less-harmful nuclear waste. In particular, the Integral Fast Reactor was a proposed nuclear
reactor with a nuclear fuel cycle that produced no transuranic waste and in
fact, could consume transuranic waste. It proceeded as far as large-scale
tests, but was then canceled by the US Government. Another approach, considered
safer but requiring more development, is to dedicate subcritical reactors to the transmutation of the left-over transuranic
elements.
An
isotope that is found in nuclear waste and that represents a concern in terms
of proliferation is Pu-239. The estimated world total of plutonium in the year
2000 was of 1,645 MT, of which 210 MT had been separated by reprocessing. The
large stock of plutonium is a result of its production inside uranium-fueled
reactors and of the reprocessing of weapons-grade plutonium during the weapons
program. An option for getting rid of this plutonium is to use it as a fuel in
a traditional Light Water Reactor (LWR). Several fuel types with differing
plutonium destruction efficiencies are under study. See Nuclear transmutation.
Transmutation
was banned in the US in April 1977 by President Carter due to the danger of
plutonium proliferation, but President Reagan rescinded the ban in 1981.[51]
Due to the economic losses and risks, construction of reprocessing plants
during this time did not resume. Due to high energy demand, work on the method
has continued in the EU.
This has resulted in a practical nuclear research reactor called Myrrha
in which transmutation is possible. Additionally, a new research program called
ACTINET has been started in the EU to make transmutation
possible on a large, industrial scale. According to President Bush's Global
Nuclear Energy Partnership (GNEP) of 2007, the US is now actively promoting
research on transmutation technologies needed to markedly reduce the problem of
nuclear waste treatment.
There have also been theoretical studies
involving the use of fusion reactors as so called "actinide
burners" where a fusion reactor plasma
such as in a tokamak,
could be "doped" with a small amount of the "minor"
transuranic atoms which would be transmuted (meaning fissioned in the actinide
case) to lighter elements upon their successive bombardment by the very high
energy neutrons produced by the fusion of deuterium and
tritium in the
reactor. A study at MIT
found that only 2 or 3 fusion reactors with parameters similar to that of the International
Thermonuclear Experimental Reactor (ITER) could transmute the entire annual
minor
actinide production from all of the light water reactors presently operating in the
United States fleet while simultaneously
generating approximately 1 gigawatt of power from each reactor.
Re-use of Waste
Main article: Nuclear reprocessing
Another
option is to find applications for the isotopes in nuclear waste so as to re-use them.
Already, caesium-137,
strontium-90
and a few other isotopes are extracted for certain industrial applications such
as food irradiation and radioisotope thermoelectric
generators. While re-use does not eliminate the need to manage
radioisotopes, it reduces the quantity of waste produced.
The
Nuclear Assisted Hydrocarbon Production Method,[55]
Canadian patent application 2,659,302, is a method for the temporary or
permanent storage of nuclear waste materials comprising the placing of waste
materials into one or more repositories or boreholes constructed into an unconventional oil formation. The thermal flux
of the waste materials fracture the formation, alters the chemical and/or
physical properties of hydrocarbon material within the subterranean formation
to allow removal of the altered material. A mixture of hydrocarbons, hydrogen,
and/or other formation fluids are produced from the formation. The
radioactivity of high-level radioactive waste affords proliferation resistance
to plutonium placed in the periphery of the repository or the deepest portion
of a borehole.
Breeder
reactors can run on U-238 and transuranic elements, which comprise the majority
of spent fuel radioactivity in the 1000-100000 year time span.
Space disposal
Space
disposal is an attractive notion because it permanently removes nuclear waste
from the environment. It has significant disadvantages, not least of which is
the potential for catastrophic failure of a launch
vehicle which would spread radioactive material into the atmosphere and
around the world. The high number of launches that would be required (because
no individual rocket would be able to carry very much of the material relative
to the total which needs to be disposed of) makes the proposal impractical (for
both economic and risk-based reasons). To further complicate matters,
international agreements on the regulation of such a program would need to be
established.
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