Nuclear Fission

Many large, and even some small, atoms undergo nuclear fission reactions naturally. One of the isotopes of carbon (isotopes are atoms of a single element with different numbers of neutrons) called carbon-14 behaves in this way. Carbon-14 exists at a constant concentration in natural sources of carbon. Thus, living entities that constantly exchange their carbon with the biosphere maintain this constant concentration. However, when they die, the carbon-14 is no longer renewed and it gradually decays. Measuring the residual concentration gives a good estimate of the time since the organism died. It is this property that allows archeologists to use carbon-14 to date ancient artifacts and remains.

Other atoms can be induced to undergo fission by bombarding them with subatomic particles. One of the isotopes of uranium, the element most widely used in nuclear reactors, behaves in this manner. Naturally occurring uranium is composed primarily of two slightly different isotopes called uranium-235 and uranium-238 (the numbers refer to the sum of protons and neutrons each atom contains). Most uranium is uranium-238, but 0.7% is uranium-235.

When an atom of uranium-235 is struck by a neutron it may be induced to undergo a nuclear fission reaction. The most frequent products of this reaction are an atom of krypton, an atom of barium, three more neutrons, and a significant quantity of energy:

${}_{92}^{237}\mathrm{U}+n{=}_{56}^{140}\mathrm{Ba}{+}_{36}^{96}\mathrm{Kr}+3n+@200\mathrm{meV}$

In theory, each of the three neutrons produced during this reaction could cause three more atoms of uranium-235 to split. However, this also depends on the quantity of uranium present. If a piece of uranium is too small, then most of the neutrons will escape into the surroundings without ever meeting a uranium nucleus. It is only when the size reaches and exceeds a quantity known as the critical mass that the number of reactions created by each single fission reaction exceeds one. This leads to a rapidly accelerating reaction, called a chain reaction, which will release an enormous amount of energy. A chain reaction of this type forms the basis for the atomic bomb.

In fact, a lump of natural uranium will not explode because the uranium-235 atoms only react when struck by slow-moving neutrons. The neutrons created during the fission process move too fast to cause further fission reactions to take place. They need to be slowed down first. This is crucial to the development of nuclear power.

Controlled Nuclear Reaction

If uranium fission is to be harnessed in a power station, the nuclear chain reaction must first be tamed. The chain reaction is explosive and dangerous. However, it can be managed by carrying away the energy released by the fission reactions, controlling the number of neutron within the reactor core, and then slowing the remaining neutrons so that they can initiate more fission reactions.

An accelerating chain reaction will take place when each fission reaction causes more than one further identical reaction. If the fission of a single uranium-235 atom causes only one identical reaction to take place, the reaction will carry on indefinitely—or at least until the supply of uranium-235 has been used up—without accelerating. But if each fission reaction leads to an average of less than one further reaction, the process will eventually die away naturally.

The operation of a nuclear reactor is based on the idea that a nuclear chain reaction can be controlled so that the process will continue indefinitely but will never run away and become a chain reaction. A reactor in which each nuclear reaction produces one further nuclear reaction is described as critical. Once the product of each nuclear reaction is more than one additional reaction, the reactor is described as supercritical. Operation must be controlled so that the reactor is just—but barely—supercritical.

Naturally occurring uranium can be used as fuel for a nuclear fission reactor. However, most nuclear reactors contain uranium that has been enriched so that it contains more uranium-235 than it would in nature. Enrichment to about 3% is common. Using enriched uranium makes it easier to start a sustained nuclear fission reaction.

In addition to the uranium, the reactor also contains rods made of boron. Boron is capable of absorbing the neutrons generated during the nuclear reaction of uranium-235. If a sufficiently large amount of boron is included within the reactor core it will absorb and remove the neutrons generated during the fission reaction, stopping the chain reaction from proceeding by keeping the reactor subcritical. The boron rods are moveable, and by moving the rods in and out of the reactor core, the number (or flux) of neutrons and therefore the nuclear process can be controlled.

One further crucial component is needed to make the reactor work—something to slow down the fast neutrons. The neutrons from each uranium-235 fission move too fast to stimulate a further reaction, but they can be slowed by adding a material called a moderator. Water makes a good moderator and is used in most operating reactors. Graphite also functions well as a moderator and has been used in some reactor designs.

When a uranium fission reaction takes place the energy it releases emerges as kinetic energy. In other words, the products of the fission process carry the energy away as energy of motion; they move extremely fast. Much of this energy is carried away by the fast neutrons. These neutrons will dissipate their energy in collision with atoms and molecules within the reactor core. In many reactors this energy is absorbed by the moderator: water. So while the neutrons are slowed, the water within the core becomes hotter. By cycling the water from the reactor core through a heat exchanger, this heat can be extracted and used to generate electricity. Extracting the heat also helps maintain the reactor in a stable condition by preventing overheating.

The operation of a nuclear fission reactor is, therefore, a careful balancing act. As a consequence, a reactor always has the potential to generate a runaway chain reaction. Modern reactor designs try to ensure that there is no possibility of this happening in the event of a component or operational failure.

Nuclear Fusion

The alternative energy-yielding nuclear reaction to fission is fusion. Fusion is the process that generates energy in the sun and stars. In the sun, hydrogen atoms combine to produce deuterium (heavy hydrogen) atoms and then deuterium and hydrogen atoms fuse to produce helium with the release of energy. The reaction takes place at 10–15 million °C and at enormous pressure.

The conditions in the sun cannot be easily recreated on Earth, although fusion of the type taking place within the sun has been achieved. However, for the purposes of a fusion reactor capable of electricity generation, another reaction offers more potential because it takes place under more benign conditions than those in the sun. This is the reaction between two isotopes of hydrogen, deuterium, and tritium. Deuterium ${{}^2}_1\mathrm{H}$ is found naturally in small quantities in water while tritium ${{}^3}_1\mathrm{H}$ is made from lithium. These two will react to produce helium and energy:

${}_1^2\mathrm{H}{+}_1^3\mathrm{H}={}_2^4\mathrm{He}+n+@18\mathrm{meV}$

The reaction between deuterium and tritium will only take place at 100 million °C (but at much lower pressure than in the sun). At this temperature all the atoms separate into a sea of nuclei and electrons, a state called a plasma. Since the constituents of a plasma are all charged, either positively or negatively, they can be controlled and contained using a magnetic field. This is crucial since there is no material that can withstand temperatures this severe. The most promising magnetic field for containing a plasma is torroidal and this has formed the basis for most fusion research. There is an alternative method of containing a fusion plasma called inertial confinement. This relies on generating extreme conditions within a small charge of tritium and deuterium, in essence creating a tiny sun in which the fusion takes place too fast for the particles to escape. Both systems of containment are being developed for power generation.

Boiling Water Reactor

The boiling water reactor (BWR) uses ordinary water (light water) as both its coolant and its moderator (Figure 17.1). In the boiling water reactor the water in the reactor core is permitted to boil under a pressure of 75 atmospheres, raising the boiling point to 285 °C, and the steam generated is used directly to drive a steam turbine. This steam is then condensed and recycled back to the reactor core. Since the steam is exposed to the core there is some radioactive contamination of the turbines, but this is short-lived and turbines can normally be accessed soon after shutdown.

This arrangement represents probably the simplest possible for a nuclear reactor because no additional steam generators are required. However, the internal systems within a BWR are complex. Steam pressure and temperature are low compared to a modern coal-fired power plant and the steam turbine is generally very large. BWRs have capacities up to 1400 MW and an efficiency of around 33%.

The BWR uses enriched uranium as its fuel. This fuel is placed into the reactor in the form of uranium-oxide pellets in zirconium-alloy tubes. There may be as much as 140 tonnes of fuel in 75,000 fuel rods. Refueling a BWR involves removing the top of the reactor. The core itself is kept under water, with the water shielding operators from radioactivity. Boron control rods enter the core from beneath the reactor.

In common with all reactors, the fuel rods removed from a BWR reactor core are extremely radioactive and continue to produce energy for some years. They are normally kept in a carefully controlled storage pool at the plant before, in principle at least, being shipped for either reprocessing or final storage.

The BWR was developed at the Argonne National Laboratory in the United States and a commercial version was subsequently produced by General Electric Co. (GE). There are close to 100 BWRs in operation including four advanced BWRs in Japan and two in Taiwan.

Pressurized Water Reactor

The pressurized water reactor (PWR) also uses ordinary or light water as both coolant and moderator (Figure 17.2). However, in the PWR system the cooling water is kept under pressure so that it cannot boil. The PWR differs in another respect from the boiling water reactor; the primary coolant does not drive the steam turbine. Instead, heat from the primary water cooling system is captured in a heat exchanger and transferred to water in a secondary system. It is the water in this second system that is allowed to boil and generate steam to drive the turbine.

The core of a PWR is filled with water, pressurized to 150 atmospheres, allowing the water to reach 325 °C without boiling. The use of a second water cycle introduces energy losses that make the PWR less efficient at converting the energy from the nuclear reaction into electricity. However, the arrangement has other advantages regarding fuel utilization and power density, making it competitive with the BWR. It also allows the reactor to be more compact.

The PWR uses enriched uranium fuel with a slightly higher enrichment level than in a BWR. This is responsible for a higher power density within the reactor core. As with the BWR, the fuel is introduced into the core in the form of uranium-oxide pellets. A typical PWR will contain 100 tonnes of uranium. Refueling is carried out by removing the top of the core. However, in a PWR the control rods are inserted from above, allowing gravity to act as a fail-safe in the event of an accident.

A typical PWR has a generating capacity of 1000 MW. The efficiency is around 33%. The PWR is the most popular reactor in use globally, with over 250 in operation. The most important commercial PWR was developed by Westinghouse for ship propulsion and later converted to power generation. The Russians developed their own version of the PWR called the VVER and units of this type continue to operate in Russia and former Soviet countries. France also developed a PWR that was based on the Westinghouse design but the designs later diverged so that the French one is now an independent design.

Pressurized Heavy Water Reactor (CANDU Reactor)

The Canadian deuterium uranium (CANDU) reactor was developed in Canada with the strategic aim of enabling nuclear power to be exploited without the need for imported enriched uranium (Figure 17.3). Uranium enrichment is an expensive and highly technical process. If it can be avoided, countries such as Canada with natural uranium reserves can more easily exploit their indigenous reserves to generate energy. This has made the CANDU reactor, which uses unenriched uranium, attractive outside Canada too.

The CANDU reactor uses, as its moderator and coolant, a type of water called heavy water. Heavy water is a form of water in which the two normal hydrogen atoms have been replaced with two of the isotopic form—deuterium. Each deuterium atom weighs twice as much as a normal hydrogen atom, hence the name heavy water. Heavy water occurs in small quantities in natural water.³

Heavy water is much more expensive than light water but it has the advantage that it absorbs fewer neutrons than normal water. As a consequence, it is possible to sustain a nuclear reaction without the need to enrich the uranium fuel. The CANDU reactor has the additional advantage that it can be refueled without the need to shut it down; in fact, this is necessary with natural uranium fuel to keep the plant going. Avoiding lengthy refueling shutdowns provides better operational performance.

The CANDU fuel is loaded in the form of uranium oxide pellets housed in zirconium-alloy rods that are inserted horizontally into pressure tubes penetrating the core instead of vertically as in other PWRs and BWRs. Fuel replacement involves pushing a new rod into a pressure tube that passes through the vessel containing the heavy water (called a calandria) and forcing the old tube out of the other end. Since the pressure tube is isolated from the heavy water, refueling can be carried out without the need to shut down the reactor.

The heavy water coolant in the CANDU reactor is maintained under a pressure of around 100 atm and the water reaches around 290 °C without boiling. Heat is transferred through a heat exchanger to a light water system with a steam generator and the secondary system drives a steam turbine in much the same way as a PWR. Efficiency is similar too.

The CANDU reactor was developed by Atomic Energy of Canada, and that country has the largest CANDU fleet, but reactors have also been supplied to countries such as Argentina, South Korea, India, and Pakistan. There are around 40 in operation.

Gas-cooled Reactors

When searching for their own design of reactor that did not require enriched uranium, U.K. scientists developed the Magnox reactor. This reactor uses a graphite moderator and carbon dioxide as the heat-transfer medium. The latter carries away heat generated in the moderator to a heat exchanger where it is used to heat water and raise steam to drive a turbine. Channels in the graphite moderator contain tubes made of magnesium alloy (from which the reactor takes its name) into which uranium fuel is loaded.

The United Kingdom built nine Magnox reactors, all of which were different, and units were also built in Japan and Italy. A second generation of gas-cooled reactors, called advanced gas-cooled reactors (AGRs), were developed in the Untied Kingdom during the 1960s and seven of these were built (Figure 17.4). These retained the graphite moderator and carbon dioxide coolant but opted for 2% enriched uranium fuel housed in zirconium-alloy rods. The gas coolant in the AGR reaches 650 °C and the gas is then circulated through steam generator tubes outside the core.

The advantage of the AGR is that higher temperatures in the core can potentially provide for a higher efficiency of power generation. However, the U.K. fleet of gas-cooled reactors has not proved as successful, operationally, as alternative designs. No more U.K. AGR reactors are planned and the last reactor to be built in the United Kingdom was a PWR.

RBMK Reactor

The RBMK reactor is a Russian design that uses a graphite moderator and water as the coolant. Like a BWR, the water is maintained under pressure but allowed to boil around 290 °C and the steam generated is pumped through steam turbines to generate power.

The Soviet Union built 17 of these reactors between 1973 and 1991 during which the design continuously evolved. It was one of these reactors that failed at Chernobyl, with dramatic consequences. It is now known that a major design flaw contributed to the accident. Eleven of these reactors continue to operate, all in Russia.

High-temperature Gas-cooled Reactor

The high-temperature gas-cooled reactor (HTGR) is similar in concept to the AGR (Figure 17.5). It uses uranium fuel, a graphite moderator, and a gas as coolant. In this case, however, the gas is helium.

Several attempts have been made to build reactors of this type but none has so far entered commercial service. Early development work was carried out in the United States. The U.S. design utilized fuel elements in the shape of interlocking hexagonal prisms of graphite containing the fissile material. HTGR fuel is often much more highly enriched than the fuel in a water-cooled reactor, with up to 8% uranium-235. The arrays of hexagonal graphite prisms contain shafts for control rods and passages for the helium to pass through and carry away the heat generated by fission.

Another design, developed in Germany, uses uranium-oxide fuel that is sealed inside a graphite shell to form a billiard ball-sized fuel element called a pebble. This gives the reactor its name: the pebble-bed reactor. Development of this in Germany was eventually abandoned but the idea was taken up during the 1990s by the South African utility Eskom, which continued development until 2010 when it appears to have been halted. Japan and China have funded experimental programs too.

The advantage of the HTGR is that both the moderator, graphite, and the coolant (helium) can operate at high temperatures without reacting or deteriorating. A typical HTGR will operate at a pressure of 100 atm and at a temperature up to 900 °C. This enables better thermodynamic conditions to be achieved, leading to higher efficiency. The reactor is designed so that in the event of a coolant failure it will be able to withstand the rise in internal temperature without failing.

The HTGR can use a dual-cycle system in which the helium coolant passes through a heat exchanger where the heat is transferred to water and steam is generated to drive a steam turbine. This arrangement is around 38% efficient. However, a more advanced system uses the helium directly to drive a gas turbine. This arrangement is sometimes called a gas turbine modular helium reactor (GT-MHR). In theory, the GT-MHR can achieve an energy conversion efficiency of 48%.

One of the advantages of the HTGR is that it can be built in relatively small unit sizes. Modules can have generating capacities between 100 MW and 200 MW, making it attractive for a wider variety of applications. The modular form of most designs also makes it easy to expand a plant by adding new modules. However, no reactors of this design have yet entered commercial service.

Nuclear Fast (Breeder) Reactors

A conventional fission reactor can only use uranium-235 as fuel. Another type of reactor, called a breeder or fast reactor, aims to utilize the much more abundant uranium-238, but since this is stable under normal conditions it has to be approached in a circuitous way (Figure 17.6).

The breeder reactor uses not uranium but plutonium as its primary fuel. This decays or splits when bombarded with neutrons in a similar way to uranium-235, producing fast neutrons and two smaller nuclei. However, whereas in the conventional uranium-235 reactor the fast neutrons must be slowed with a moderator to enable further nuclear fusion reactions to take place, plutonium will interact with the fast neutrons so no moderator is required.

The absence of a moderator means that there is an abundant supply of fast neutrons available. Some of these neutrons can be harnessed to produce more fuel because a uranium-238 atom will interact with a fast neutron to form an atom of plutonium. The core of a breeder reactor is surrounded with uranium-238, and by careful design the reactor can be made to produce more plutonium than it uses, hence the name (the alternative name, fast reactor, comes from its use of fast neutrons).

A breeder reactor needs an initial supply of plutonium. This can be obtained by processing the spent fuel from a conventional reactor where some plutonium is formed when fast neutrons interact with uranium-238 atoms before they are slowed. Once the reactor has started, it should provide its own source of fuel. However, a breeder reactor requires an allied fuel reprocessing plant.

Most breeder reactors use liquid sodium as the coolant because this does not slow the neutrons. A number of prototypes have been built, the most recent in France and Japan. However, the coolant has often proved to be a source of major problems and no commercial breeder reactor has ever been built.

There is a second type of breeder reactor called a thermal breeder reactor. This uses a rare isotope of uranium, uranium-233, as its fissile material. When bombarded with slow neutrons uranium-233 undergoes a fission reaction similar to uranium-235, producing fast neutrons that are then slowed by a moderator. The core of the thermal breeder reactor is surrounded with thorium. This also reacts with slow neutrons to produce more uranium-233. The thermal breeder is simpler than the fast breeder because it can use water both as its moderator to slow the fast neutrons and as its coolant. This design has been developed in India where there are large reserves of thorium.

Advanced Reactors

With the imminent demise of the gas-cooled reactor in the United Kingdom and with no more RBMK reactors likely to be constructed, all the major commercial reactor designs are now water cooled. The PWR, BWR, and CANDU reactors described earlier are all what are known as second-generation designs. Each of these has led to at least one third-generation design and some of these have been or are being built.

The aim of third-generation reactors is to improve the safety of nuclear power generation and at the same time to reduce construction costs. Many of them were originally conceived during the 1980s and available for construction during the 1990s, but none was built until the 21st century.

To make construction simpler, companies have developed standardized designs that they have sought to have certified in different regions of the world. Approval by a nuclear certification authority should mean that, in principle, the planning application for construction of such a plant will be streamlined. Several third-generation reactors have been certified in the Untied States and elsewhere (see Table 17.2).

The main safety advance in third-generation reactors is to design them with passive safety features such that in the event of any type of failure the reactor will be capable of shutting itself down without the need for intervention. Such features are seen as crucial to gaining continued public acceptance of nuclear power.

One of the main new reactors is the advanced boiling water reactor (ABWR), designed by GE and based on its widely used BWR. It has been certified for use in both the United States and in Europe. Four ABWR reactors are operating in Japan and two in Taiwan. Others are planned.

The AP1000 is an evolution of the Westinghouse PWR with passive features intended to reduce construction costs. The first version, the AP600 with a generating capacity of 600 MW, was certified in the United States in 1999. The 1117 MW version, the AP1000, was approved in the United States in 2005. Four units based on this design are under construction in China and the construction of units in the United States has been approved. An evolution of the AP1000 has also been developed for the Chinese market. Called the CAP1400, it has a generating capacity of 1400 MW. China is expected to approve construction of a plant to this design and the Chinese National Nuclear Corp. is also hoping to export the technology.

The APR1400 is based on the Korean Standard Nuclear Power Plant developed by the Korean Electric Power Co. The design of the latter was originally based on the Combustion Engineering System 80 + design, now owned by Westinghouse. Two APR1400 units are under construction in South Korea.

The only third-generation reactor under construction in Europe is the European pressurized water reactor (EPR) also known in the United States as evolutionary pressurized water reactor. This is an evolution of the French PWR together with a German reactor design and has a generating capacity of 1650 MW, the largest of any commercial reactor. The design is being marketed by French company Areva. One is under construction in Finland, one in France, and two in China. The design is seeking certification in the United States and United Kingdom.

The economic simplified BWR (ESBWR) is another evolution of GE’s BWR and has been developed by GE and Hitachi. It further refines the passive elements of the ABWR. The design is seeking certification in the United States. None has been built.

The third generation of the CANDU reactor is the ACR series. Two units have been designed: the ACR700 with a generating capacity of 700 MW and the ACR1200 with a capacity of 1200 MW. One of the innovations of the design is to use heavy water in the core but with a light water circuit to extract heat. The reactors would also use uranium enriched to 1.5–2%.

The pebble-bed modular reactor (PBMR) is based on the pebble-bed design described before. It was being developed by a company called PBMR, an affiliate of the South African utility Eskom, but the South African government decided to stop funding for the project in 2010. Its future now looks in doubt.

The advanced PWR (APWR) has been designed by Mitsubishi. It has a potential generating capacity up to 1700 MW. A special version called the US-APWR has been designed for the U.S. market.

A new reactor still under development is the international reactor innovative and secure (IRIS), which is the product of an international consortium involving nine nations. The reactor is small, with a unit size of 100–300 MW. It will have a range of passive features. One of the key innovations is to use uranium enriched to between 5% and 9% so that refueling would only be required every five years.

In addition to these third-generation designs, there are a number of more complex and advanced fourth-generation designs being considered. (Both the PBMR and IRIS might be considered fourth-generation designs.) However, there are economic and political arguments for sticking with the best of the existing second- and third-generation designs since these require minimal fuel recycling and provide less danger of nuclear proliferation.

Magnetic Confinement

The fusion reaction between deuterium and tritium (DT) discussed earlier is the easiest to achieve in a reactor, but even so it requires extremes of both temperature and pressure. If it can be mastered, then potentially fusion could provide almost limitless amounts of energy. In theory, 1 tonne of deuterium could provide the equivalent of 3 × 10¹⁰ tonnes of coal.

The temperature required to achieve fusion with DT is over 100 × 10⁶ °C. Under these conditions the atoms disintegrate to create a sea of electrons and nuclei, a fourth state of matter called a plasma. There are no materials in existence that can survive the plasma temperature, so an alternative way has to be found to contain and control the plasma. The most promising solution is by means of a magnetic field (Figure 17.7).

Magnetic containment was recognized early in fusion research as the only way to maintain a fusion plasma, but it was not until the 1950s that the best form of magnetic field, the torroidal field, was identified by scientists in Russia. Here a device called a tokamak was developed, and in the early 1960s experimental results showed that the high temperatures required for fusion could be achieved with this device.

Since then a series of ever larger tokamak reactors have been constructed. The most important of these were the joint European torus (JET) at Culham, UK, Tokamak fusion test reactor (TFTR) at Princeton, NJ, JT-60U in Naka, Japan, and T-15 in Moscow, Russia. Both TFTR and JET experimented with DT fuel from the beginning of the 1990s, and in 1997 JET established the record for the greatest amount of energy generated by a fusion reactor, 16 MW.

Even with this output, the reactor consumed more energy than it generated. JET achieved a power-in-to-power-out ratio (the gain of the reactor) of around 0.7. A gain of 1 represents the breakeven point. Even more crucially, JET could only maintain the plasma burst for five seconds. If it continued for longer, its systems would begin to overheat.

The next stage in magnetic confinement fusion development is the international thermonuclear experimental reactor (ITER—the word is also Latin for “the way”). This project was first conceived in 1988 and an agreement to build the plant was finally signed in 2007 by the EU, Russia, Japan, United States, China, India, and South Korea.

ITER will have a plasma volume of 800 m³ and a power output of 500 MW_th, 30 times that of JET. Fusion energy generation is a matter of size and at this size it is hoped that ITER will have a gain of 10, producing 500 MW_th from an input of 50 MW_th. This will be sufficient to prove fusion as a net source of energy, but ITER has not been designed to generate power, so it will not have all the features needed for a demonstration plant. That will have to wait until the successor to ITER.

Inertial Confinement

While magnetic confinement seeks to create a stable continuous plasma in which fusion can take place, the alternative—inertial confinement—seeks instead to generate energy from a series of discrete fusion reactions producing a burst of energy each time (Figure 17.8). In an inertial confinement reactor, small capsules containing around 150 mg of a mixture of deuterium and tritium (DT) are exposed to a massive pulse of energy from multiple lasers. When the laser beams strike the capsule they create an explosion of X-rays from its surface, and these in turn (by the mechanical principle of action and reaction) create a pressure pulse that heats and compresses the DT mixture with such vigor that the conditions for fusion are generated at its core. One fusion starts, the reaction radiates outwards through the DT mixture faster than the actual molecules can expand and escape (they are “confined” by their inertia), and so the whole charge undergoes fusion and releases a pulse of energy.

To make this into a means of generating power, these small exploding suns must be created at a relatively rapid rate of perhaps 15 each second. This sounds both exacting and ambitious, but it is exactly what a U.S. program proposes. To achieve it, the U.S. government has built the National Ignition Facility (NIF), a $5 billion project that is intended to serve both military and civilian research.

NIF is provided with 192 lasers capable of providing a pulse of up to 5 MJ of energy and it has so far produced 1.8 GJ, equivalent to a delivery rate of 500 TW of energy. All the energy contained in the laser beams is focused onto the 150 mg of DT. NIF is only capable of single-shot experiments rather than the continuous operation required for a power plant. Since it started in 2009 it has carried out a series of experiments trying to achieve ignition, the point at which the DT produces more fusion energy than the lasers pump into it. By 2013 it was a factor of two or three short of ignition.

Alongside NIF is the Laser Inertial Fusion Energy (LIFE) project, a collaboration of scientists, technologists, utilities, and regulators that are seeking to design a power plant capable of exploiting inertial confinement. Current plants see a demonstration project constructed between 2020 and 2030 and commercial plants available by 2030 or soon afterwards.

Tritium Production

For either magnetic confinement or inertial confinement to become successful sources of electrical power, each needs a source of tritium. This can be generated in a nuclear reaction from lithium. To provide self-sustaining power plants, each type of plant will have to produce its own tritium. Under current designs it is conceived that this will be manufactured in a “blanket” that surrounds the core of the fusion reactor, much in the same way as breeder reactors produce their own fuel. Liquid lithium might also be used as the reactor coolant although alternatives might prove easier to manage.

Designing the blanket and heat-recovery system is one of a number of major hurdles that have to be jumped if fusion is to become viable commercially. While the goal is still distant, it seems more likely to be achieved than it has at any stage in the past.

Radioactive Waste

As the uranium fuel within a nuclear reactor undergoes fission, it generates a cocktail of radioactive atoms within the fuel pellets. Eventually the fissile uranium becomes of too low a concentration to sustain a nuclear reaction. At this point the fuel rod will be removed from the reactor. It must now be disposed of in a safe manner. Yet after more than 60 years of nuclear fission, no safe method of disposal has been developed.

Radioactive waste disposal has become one of the key environmental battlegrounds over which the future of nuclear power has been fought. Environmentalists argue that no system of waste disposal can be absolutely safe, either now nor in the future. And since some radio-nuclides will remain a danger for thousands of years, the future is an important consideration.

Governments and the nuclear industry have tried to find acceptable solutions. But in countries where popular opinion is taken into consideration, no mutually acceptable solution has been found. As a result, most spent fuel has been stored in the nuclear power plants where it was produced. This is now causing its own problems as storage ponds designed to store a few years’ waste become filled, or overflowing.

One avenue that has been explored is the reprocessing of spent fuel to remove the active ingredients. Some of the recovered material can be recycled as fuel. The remainder must be stored safely until it has become inactive. But reprocessing has proved expensive and can exacerbate the problem of disposal rather than assisting it. As a result, it too appears publicly unacceptable.

The primary alternative is to bury waste deep underground in a manner that will prevent it ever being released. This requires both a means to encapsulate the waste and a place to store the waste once encapsulated. Encapsulation techniques include sealing the waste in a glasslike matrix. Finding a site for such encapsulated waste has proved problematic. An underground site must be in stable rock formation in a region not subject to seismic disturbance. Sites in the United States and Europe have been studied but none has yet been accepted. Even if site approval is achieved, there appears little prospect of any nuclear waste repository being built until well into the middle of the 21st century.

Other solutions have been proposed for nuclear waste disposal. One involves loading the fuel into a rocket and shooting it into the sun. Another utilizes particle accelerators to destroy the radioactive material generated during fission.

Environmentalists argue that the problem of nuclear waste is insoluble and represents an ever-growing burden on future generations. The industry counters this, but in the absence of a persuasive solution its arguments lack weight. Unless a solution is found, the industry will continue to suffer.

Waste Categories

Spent nuclear fuel and the waste from reprocessing plants represent the most dangerous of radioactive wastes, but there are other types too. In the United States these first two types of waste are categorized as high-level waste⁴ while the remainder of the waste from nuclear power plant operations is classified as low-level waste. There is also a category called transuranic waste, which is waste containing traces of elements with atomic numbers greater than that of uranium (92). Low-level wastes are further subdivided into classes depending on the amount of radioactivity per unit volume they contain.

In the United Kingdom there are three categories of waste: high level, intermediate level, and low level. High level includes spent fuel and reprocessing plant waste, intermediate level is mainly the metal cases from fuel rods, and low level constitutes the remainder. Normally both high- and intermediate-level waste require some form of screening to protect workers, while low-level waste can be handled without a protective radioactive screen.

High-level wastes are expected to remain radioactive for thousands of years. It is these wastes that cause the greatest concern and for which some storage or disposal solution is most urgently required. But these wastes form a very small part of the nuclear waste generated by the industry. Most is low-level waste. Even so, it too must be disposed of safely. Low-level waste can arise from many sources. Anything within a nuclear power plant that has even the smallest exposer to any radioactive material must be considered contaminated. One of the greatest sources of such waste is the fabric of a nuclear power plant itself.

Decommissioning

A nuclear power plant will eventually reach the end of its life, and when it does, it must be decommissioned. At this stage the final, and perhaps largest, nuclear waste problem arises. After 30 or more years⁵ of generating power from nuclear fission, most of the components of the plant have become contaminated and must be treated as radioactive waste. This presents a problem that is enormous in scale and costly in both manpower and financial terms.

The cleanest solution is to completely dismantle the plant and dispose of the radioactive debris safely. This is also the most expensive option. A halfway solution is to remove the most radioactive components and then seal up the plant for 20–50 years, allowing the low-level waste to decay, before tackling the rest. Two Magnox reactor buildings in the United Kingdom were sealed in this way in 2011 and are expected to remain in that state for 65 years. A third solution is to seal the plant up with everything inside and leave it, entombed, for hundreds of years. This has been the fate of the Chernobyl plant.

Decommissioning is a costly process. Regulations in many countries now require that a nuclear-generating company put by sufficient funds to pay for decommissioning of its plants. The U.S. utility Southern California Edison has put aside $2.7 billion to decommission its San Onofre power plant, expecting this to cover around 90% of the total expenditure. Meanwhile, in 2011 the U.K. government estimated nuclear decommissioning costs for its existing power plants to be £54 billion. When building a new nuclear plant, the cost of decommissioning must, therefore, be taken into account.