6.1 Introduction

Nuclear power stations are used in thirty-one countries or 15% of the total number of countries in the world. Moreover, the generated nuclear power covers 13% of world energy consumption (WNA, 2013), which represents ~ 2.5 PW. Currently, the most nuclear power is generated in the United States with a 31.3% share, followed by France with a 16.8% share; some other countries with an important share of nuclear power generation are the Russian Federation (6.4%), Japan (6.2%), South Korea (5.8%), Germany (4%), Canada (3.5%), Ukraine (3.4%), and China (3.3%).

All nuclear power stations worldwide use uranium as nuclear fuel extracted from uranite mineral resources. The richest country in uranite is Australia, with a 23% share, followed by Kazakhstan with 15%, Russia with 10%, South Africa and Canada with 8%, United States of America with 6%, Brazil, Namibia, and Niger with 5%, Ukraine with 4%, Jordan and Uzbekistan with 2%, and India with 1%.

In a nuclear power station a controlled nuclear reaction is maintained in the reactor. The nuclear reaction generates mainly neutron radiation that is immediately converted to high-temperature heat. A steam generator with special design—depending on the reactor type—is then used to extract the thermal energy produced by the reaction and to obtain pressurized superheated steam at 525–625 K. The steam generator is part of a steam Rankine power plant which produces power with a typical energy efficiency of 30%.

The key to achieving the current level of technology of nuclear power stations has been the establishment of a new body of science in the last quarter of the nineteenth and the first quarter of the twentieth centuries: nuclear physics. The first laboratory demonstration proving the possibility of generating nuclear radiation in a controlled manner was realized in 1924 by the research group of the well-known scientists Frederic and Irene Joliot Curie in Paris and their PhD student Stefania Maracineanu, which showed that lead, being activated with radioactive polonium, starts emitting radiation (Neues Wiener Journal, 1934). Further experiments of the research group led to the formulation of a theoretical model for artificial radioactivity in aluminum bombarded with alpha particles, a discovery which was of utmost importance for the progress of nuclear energy engineering; this discovery was acknowledged by the 1935 Nobel Prize for chemistry awarded to Frederic and Irene Joliot Curie.

The proof of the concept for a nuclear reactor was achieved only 7 years later in Chicago by Enrico Fermi, who demonstrated the first nuclear reactor with a controlled chained fission reaction of uranium. Later a nuclear reactor of 1 MW was built and tested at Oak Ridge Laboratories (1943) and the first large-size nuclear reactor (of 250 MW installed capacity), called the B reactor, began to operate near Richland Village, WA (1944). The first nuclear power plant was commissioned in Idaho, USA, in 1951, with an installed electricity of 100 kW. CANada Deuterium Uranium (CANDU) technology achieved a commercial level in 1962. During the last 40 years, many kinds of nuclear reactors have been developed for the purpose of electric power production and marine and space mission applications.

Nuclear power generation is a crucial player in the generation of wealth, economic development, and energy security. In this chapter, nuclear power generation is discussed starting from the most basic concepts of nuclear reaction, to the engineering of controlled generation of nuclear heat and power cycles, to the advanced concepts of a nuclear reactor. As cogeneration brings a better recognition and wider acceptance of nuclear energy for daily applications in various sectors, this topic is also discussed, with emphasis on producing district heating, process heat, potable water from desalination, hydrogen, and other synthetic fuels.

6.2 Nuclear Reactions

It is well known that some chemical elements are radioactive; they can transform themselves into other chemical elements and emit or absorb radiation through nuclear processes such as radioactive decay, or they can participate in fusion or fission reactions. How do nuclear reactions occur and how can they be controlled for steady generation of high-temperature heat? This is the main question that we want to answer in this chapter, by providing the basic elements of nuclear energy and power generation applications.

Let us recall that the atom is the smallest unit of a chemical element (radius of about 1 ) and that there are 118 known atoms corresponding exactly to the periodic table of elements. The atom is made up of a nucleus and a number of electrons. The nucleus is formed of protons and neutrons. In normal condition the atom is neutral with respect to electric charge; thus, the number of electrons is the same as the number of protons. The nucleus cannot be broken down by chemical reactions, but it can be affected in its constituency by nuclear reactions. Table 6.1 gives the formal definition of the elementary particles which are considered the most significant ones for nuclear power generation.

Table 6.1

A List of Significant Particles in Nuclear Physics

Particle	Symbol	Definition
Electron	e−	Subatomic particle having one negative elementary electric charge and a very small rest mass, negligible with respect to the nucleus mass
Positron	e+	The antimatter that annihilates an electron having the same rest mass but positive charge
Proton	p+	Subatomic particle having one positive elementary electric charge and rest mass. Proton is categorized as a nuclide because it enters into the constituency of the atomic nucleus
Neutron	n	Subatomic particle having no electric charge and a rest mass approximately equal to that of proton. The neutron is categorized as a nuclide
Neutrino	ν	This particle has zero electric charge and non-zero rest mass; it is smaller than an electron and travels with a speed close to that of light
Antineutrino	$\overline{\nu }$	The antimatter that can annihilate a neutrino
Photon	hν	The elementary quanta of electromagnetic radiation (light). Photon has a dualistic behavior as wave and particle. Its speed of propagation is the speed of light; therefore it has no rest mass. It has no electric charge, but it has dynamic mass, momentum, and spin. Photons can be annihilated by absorption when light interacts with matter

The nucleus has a rest mass in the range of 10^− 24–10^− 22 g. In some specific nuclear reactions, a portion of the nucleus mass may disappear, as it can be converted into radiation energy according to the well-known Einstein formula, which is written here ΔE = Δm c² to emphasize that a change of nucleus mass of Δm corresponds to an emitted (or absorbed) energy amount noted with ΔE. In order to have an idea of the energy generated by nuclear reactions, let’s assume that a mass of 10^− 25 g is lost by a nucleus which represents the thousandth part of the average atomic mass of the chemical elements. The generated energy caused by the conversion of this amount of mass into radiation is calculated to be 56 MeV. If 5 kmol of atoms (~ 300 kg radioactive matter, on average) would react in the same way, then the energy released is 27 EJ, or the same as the annual nuclear power generated worldwide with the installed capacity; this gives an idea of the immense potential of using nuclear reactions for power generation.

The balance of atomic masses is a very important issue for analysis of nuclear reactions. At the same time, the atomic mass (denoted with A, also known as mass number) is a parameter that identifies chemical elements. Any chemical element is identified by the number of protons in its nucleus, called the atomic number (or proton number) and denoted with Z. However, for a chemical element the number of neutrons may differ from the number of protons within the nucleus. A chemical element may have some degree of variety in its nucleus corresponding to a variation in the number of neutrons.

Therefore, although the electric charge of the nucleus of a chemical element is fixed, the nucleus mass varies depending on the number of neutrons generating a range of subspecies. Nuclei having the same number of protons and a different number of neutrons are called isotopes. The abundance of an isotope represents the probability of identifying the particular isotope in nature among all known isotopes of the same chemical element.

Note that one isotope of particular interest is carbon-12 (₆¹²C), which is the most abundant isotope of carbon in the universe with an abundance of 98.89%. This isotope has Z = 6 protons and N = 6 neutrons and has a mass of 1.992 × 10^− 26 kg. By convention, ₆¹²C is used to define the atomic mass unit (AMU) according to the following statement:

Carbon-12 has 12 AMU.

One can simply calculate 1 AMU = 1.992 × 10^− 26/12 = 1.660538782 × 10^− 27kg. In Table 6.2, the abundance and the atomic masses of some selected isotopes relevant to nuclear-power-generating reactors are given. There are a number of universal constants that characterize the subatomic processes. These constants and other relevant constants for nuclear processes are given in Table 6.3. Nuclear reactions result in emission or absorption of various types of nuclear radiation. According to quantum mechanics, the matter absorbs and emits radiative energy in fixed packages (quanta). In general, nuclear radiation consists of a flux of unstable particles. The basic types of radiation encountered in nuclear reactions are listed and described in Table 6.4.

The release and absorption of energy during the nuclear reaction can be determined based on the conservation of energy principle which in quantum mechanics must account for the possibility of mass-to-energy conversion. The reconfiguration of the nucleus after the reaction is responsible for the released or absorbed energy. Both attractive and repulsive forces manifest among nuclides that form the nucleus. Coulomb forces produce proton repulsion and the Yukawa forces attract the nuclides. The equilibrium between attractive and repulsive nuclear forces can be stable, unstable, or metastable depending on the exact situation. If the equilibrium in the nucleus is stable, then the isotope is stable; if the nuclear equilibrium is unstable or metastable, then the isotope is radioactive—that is, it will spontaneously disintegrate to form a system with stable equilibrium.

One can determine the binding energy for any nucleus if one assumes that the nucleus discomposes in its protons and neutrons. Consider an element E with Z protons and A nuclides (A is also denoted as the atomic number). The reaction _Z^AE → Zp⁺ + (A − Z)n + ΔE_b describes the un-binding process of the nucleus with the release of the binding energy  ΔE_b. This energy is caused by the difference in mass between the protons and neutrons forming the nucleus and the mass of nucleus, which is heavier. Therefore, the binding energy of any nucleus can be calculated based on ΔE_b = c² Δm according to the formula:

$\mathrm{\Delta }{E}_{\mathrm{b}}=-\left[Z\times m_{{\mathrm{p}}^{+}}+\left(A-Z\right)m_{\mathrm{n}}-m\left({}_Z{}^{A}E\right)\right]c^2$

Note that for a mass change of 1 AMU the binding energy is ΔE_b1 = 931.4 MeV. A factor that quantifies the nuclear stability of an isotope is given by the ratio of binding energy to the atomic mass, denoted with the nucleus stability factor (NSF)

$\mathrm{NSF}=\frac{\mathrm{\Delta }{E}_{\mathrm{b}}}{m_{\mathrm{nucleus}}}\left[\frac{\mathrm{MeV}}{\mathrm{AMU}}\right]$

Because the mass of proton and the mass of neutron as well are both of the order of ~ 1 AMU (see Table 6.3) it results that the mass of nucleus is approximately equal to the sum of protons and neutrons. Therefore, the NSF gives the binding energy per number of nuclides. When the binding energy per number of nuclides is higher the nucleus is more stable. In Figure 6.1 one can observe that ⁵⁶Fe has about the maximum binding energy per nucleon (the peak is very close to ⁵⁶Fe and corresponds to less abundant isotopes of ⁶²Ni followed by ⁵⁸Fe). The nucleus of ⁴He is also very stable, showing a peak with respect to its neighbors. This high stability is the reason this isotope results from many nuclear reactions, including fusion that is believed to occur in the sun’s core, where hydrogen is converted into ⁴He, or the alpha decay where a fast moving nucleus of ⁴He is generated.

Nuclei that have a small atomic mass (smaller than that of magnesium, atomic number Z = 24) tend to associate and form heavier nuclei through nuclear-fusion-type reactions. Atoms naturally repel each other, so fusion is easiest with these lightest atoms. However, to force the atoms together takes extreme pressure and temperature.

Elements with atomic mass in the range between magnesium and xenon (atomic number Z = 131) are relatively stable as the stability factor is relatively flat. For elements heavier than xenon the nucleus size becomes large, a fact that affects the balance of forces: attractive forces become weaker and the repulsive forces stronger. Therefore, heavy elements have the tendency to disintegrate into lighter elements through a nuclear fission reaction. The most easily fissionable elements are the isotopes are uranium-235 and plutonium-239.

Fissionable elements are flooded with neutrons causing the elements to split. When these radioactive isotopes split, they form new radioactive chemicals and release extra neutrons that create a chain reaction if other fissionable material is present. While uranium (atomic number 92), is the heaviest naturally occurring element, many other elements can be made by adding protons and neutrons to nuclei with the help of particle accelerators or nuclear reactors (Figure 6.1).

Besides fusion and fission reactions there is another nuclear process specific to unstable nuclei, namely the radioactive decay. The non-stable isotopes are denoted as radionuclides. In a radioactive decay a radionuclide discomposes by emitting nuclear radiation (alpha, beta, or gamma). Assume that a batch of matter contains N₀ radionuclide at an initial moment. Due to decay reaction the number of radionuclides reduces after time t to N(t). The decay law expresses the time required for the number of radionuclides to reduce to half. The half-life time t_1/2 is defined according to the following equation:

$\frac{N\left(t\right)}{N_0}={\left(\frac{1}{2}\right)}^{t/t_{1/2}}$

The half-life time of some radionuclides are given in Table 6.5. The half-life time is a very important parameter for designing nuclear fuel cycles. Nuclear fuels after use in a nuclear reactor are still radioactive in some measure. The half-life time allows for predicting the radiation decay of spent fuel in fuels repositories. The nuclear fuel cycle is detailed in the next section.

Here we will focus on two main nuclear reactions that are relevant for power generation: thermonuclear fusion and fission. A basic reaction for thermonuclear fusion is

${}_1{}^2\mathrm{H}+{}_1{}^2\mathrm{H}\to {}_2{}^4\mathrm{H}\mathrm{e}+28.4\mathrm{MeV}$

where the thermal energy emitted by the reaction is mainly gamma radiation. The most difficult problem with this reaction consists of making two deuterium nuclei collide.

There is considerable activation energy of ~ 0.02 MeV per collision required by reaction (6.2). However, the activation energy is negligible with respect to the amount of generated energy, which is 28.4 MeV for each collision. In order to transfer the activation energy for reaction initiation the temperature of the deuterons (deuterium nuclei) must be raised to 144 × 10⁶ K, which is probably possible on the sun but definitely not on earth. Another thermonuclear reaction, practically demonstrated on earth through an explosive process, is the fusion of deuteron with triton (the tritium nucleus) according to

${}_1{}^2\mathrm{H}+{}_1{}^3\mathrm{H}\to {}_2{}^4\mathrm{H}\mathrm{e}+n+17.6\mathrm{MeV}$

The activation energy for reaction (6.3) is ~ 10 keV, with an associated temperature of 77 × 10⁶ K. In an explosive process (the hydrogen bomb) this temperature is achieved via a pilot reaction of uranium-235 fission. However, with current technology steady, controlled generation of thermal energy from the fusion reaction is not possible, especially because the process has an explosive nature with an extremely short duration of 1 μs. Research is underway to develop power generation technology from nuclear fission using various options such as electromagnetic fields and plasma and thermally induced shock waves to maintain the reaction confinement space at high temperature.

Nevertheless the most important reaction for nuclear power technology is currently the nuclear fission of uranium, a naturally occurring fuel resource. In a fission reaction a heavy nucleus is “bombarded” with a “slow” neutron. By capturing a neutron the nucleus enters into a metastable state. In order to reach equilibrium, the nucleus splits in two parts of about half weight and generates neutron radiation and energy.

The only known naturally occurring fissionable isotope is ₉₂²³⁵U. However, the abundance of this isotope on earth is only 0.7%. Only two other fissionable artificial isotopes are known: the uranium isotope ₉₂²³³U and the plutonium isotope ₉₄²³⁹Pu. Both these isotopes can be produced starting from more abundant resources through neutron bombardment. Plutonium-239 is obtained from ₉₂²³⁸U (~ 99.3% natural occurrence) through the following overall reaction

${}_{92}{}^{238}\mathrm{U}+n\to {}_{94}{}^{239}\mathrm{P}\mathrm{u}+2{\beta }^{-}$

The artificial fissionable uranium isotope (uranium-233) is produced from thorium-232 (abundance ~ 100%) through the overall reaction

${}_{90}{}^{232}\mathrm{T}\mathrm{h}+n\to {}_{92}{}^{233}\mathrm{U}+2{\beta }^{-}$

The nuclear fission process of uranium-235 requires one slow thermal neutron n_th and generates a number κ > 1 of fast neutrons and heat (in the form of gamma radiation). The first step of the reaction is the bombardment of a nucleus of ₉₂²³⁵U with one thermal neutron. As an intermediary product the process generates the metastable uranium-236, ₉₂^236mU. The fission reaction can be written in a simplified manner as follows:

${}_{92}{}^{235}\mathrm{U}+n_{\mathrm{th}}\to {}_{92}{}^{236\mathrm{m}}\mathrm{U}\to {}_{Z_1}{}^{A_1}E_1+{}_{92-Z_1}{}^{236-\kappa -A_1}E_2+\mathit{\kappa n}+Q_{\mathrm{release}}$

The metastable ₉₂^236mU is formed, which further splits into two lighter nuclei, namely ${}_{Z_1}{}^{A_1}E_1$ and ${}_{92-Z_1}{}^{236-\kappa -A_1}E_2$, and emits κ neutrons. The mass number A₁ (i.e., the number of nucleons) of the first element $\left({}_{Z_1}{}^{A_1}E_1\right)$ can be found in the range of 75–160, with a higher occurrence between 92 and 144. The resulting elements E₁ and E₂ have metastable nuclei and therefore will break apart, emitting gamma rays.

For reaction (6.4) the mean energy released by one mole of fissionable nuclear fuel (any of ₉₂²³⁵U, ₉₂²³³U, ₉₄²³⁹Pu) is Q_released = 200 MeV = 19.3 × 10¹² J. With this, one calculates 82.13 TJ per kg of ₉₂²³⁵U, 82.83 TJ per kg of ₉₂²³³U, and 80.75 TJ per kg of ₉₄²³⁹Pu. For natural uranium fuel, accounting for the occurrence of 0.7%, the associated exergy is ~ 584 GJ/kg.

6.3 Nuclear Fuel

As mentioned in the introduction the fuel for nuclear power stations is extracted from uranium ore—uranite—which may contain up to 0.3% uranium oxide. Most uranite resources are found in Australia, including 23% of the global share. The map shown in Figure 6.2 illustrates the worldwide distribution of uranium fuel resources. The total known resources are on the order of 5 Tg (teragrams, 10¹² g), with estimated theoretical resources (not yet proven) of 16 Tg. In addition to uranium there are about 20 Tt resources of fertile Thorium-232 fuel.

Once the fuel is extracted from natural deposits it follows a specific fuel processing chain until it is fed to the nuclear reactors. Practically all primary nuclear material used throughout the world is uranium, even though some limited applications may use thorium. The worldwide consumption of uranium fuel for power generation is represented on the map in Figure 6.3. For the future, thorium-232 is foreseen as the principal material to be extracted by mining for producing nuclear fuel.

After use at nuclear power stations the spent fuel—still radioactive—is further processed and eventually disposed. Various types of repositories for nuclear spent fuel exist, as will be presented later in this section, when the back-end of the fuel cycle will be discussed.

The nuclear fuel cycle comprises a front-end phase, a fuel utilization phase, and a back-end phase. The front-end covers all the processing, starting with mining and ending with fuel fabrication. The back-end includes fuel recycling, materials recovery, and disposal.

Figure 6.4 shows the main processing steps of a uranium fuel cycle and gives the approximate energy balances for producing 100 Mt of reactor fuel. At the mining site the uranium mineral is mined, milled, and extracted using chemical leaching. The resulting product is the uranium cake, which is the rough commercialized form of uranium fuel containing triuranium octaoxide (U₃O₈). About 260 Gt of ore is required to eventually produce 0.1 Gt of fuel. If enrichment is needed then U₃O₈ is converted to UF₆ and processed through isotope separation to increase the amount of ²³⁵U to 3.5%. If enrichment is not needed the triuranium octaoxide is converted into uranium dioxide (UO₂) which can be fed directly to reactors (e.g., CANDU). At the extraction phase of triuranium octaoxide about 520 Mt of product is generated for 100 Mt of fuel while close to 260 Gt of mill tailings are deposited.

The conversion of U₃O₈ to UO₂ is through a thermal process (heating is required). If enrichment is required the uranium dioxide must be further converted to uranium hexafluoride. The steps for the entire conversion process from U₃O₈ to UO₂ are illustrated in Figure 6.5. After uranium dioxide is obtained a hydro-fluorination process follows in which uranium dioxide is reacted with anhydrous hydrofluoric acid (HF). In a further conversion step the uranium tetrafluoride is combined with fluorine in gas phase to obtain uranium hexafluoride. The product extraction is done through distillation and crystallization using special steel drums. From 770 Mt of UF₆ only 150 Mt of enriched fuel can be obtained; the residual material is stored for future use with the next generation of breeder reactor.

Further processing includes molding the uranium-fuel into a pellet form. There are various ways in which the fuel pellets can be assembled and clad. Recycled fertile materials are also embedded in fuel pellets (e.g., plutonium and depleted uranium). In commercial reactors for power generation the fuel is used in solid form. Liquid fuels also exist for advanced applications and research, including liquid uranyl salts and molten salts.

A typical fuel element is sintered and formed into cylindrical cartridges with a core of uranium which can be of a metallic material (the uranium), of an alloy (e.g., with aluminum), or of uranium dioxide (UO₂) or uranium carbide (UC). One important aspect is fuel cladding. Nuclear fuel must be embedded in a layer of corrosion resistant material that allows for good heat transfer between the fuel core and the reactor coolant. Cladding has a reduced cross-section for the absorption of thermal neutrons and at the same time does not allow escape of radioactive materials. The cladding is made normally in zirconium alloys (2.5% niobium and 97.5% zirconium, which is highly penetrative to neutrons) or, as in more recent technologies, in silicon carbide. One of the problems with fuel cladding is its catalytic effect of splitting water molecules and hydrogen production; this effect is more pronounced in case of accidents when the temperature of the cladding increases and the catalytic activity enhances. Silicon carbide cladding shows very reduced water-splitting activity.

Recent advances in nuclear fuel manufacturing led to the development of tristructural isotropic-coated fuel particles (TRISO) which have improved safety and versatility. These fuel particles are designed to retain fission fragments under any possible circumstances, as they have a strong negative coefficient of reactivity. This means that TRISO fuel is intrinsically safe because there is no need for active cooling in case of accident and the reaction stops immediately when the neutron rate is reduced.

TRISO comprises a 0.5 mm spherical fuel particle of uranium dioxide which is coated in four protective shells, as shown in Figure 6.6. The outer layer acts a pressure vessel, not allowing the passage of any radioactive particles up to temperatures as high as 1900 K. TRISO particles have a diameter of 1 mm and can be formed as pebbles (of ~ 6 cm in diameter) or as hollow cylinders of 25 mm in diameter and 4 mm in length.

After fuel is used, it is discharged from the reactor and stored in interim locations. Further, the used uranium is transported to disposal places for permanent storage, or reused, depending on the case. For breeding reactors, the uranium fuel is always recycled. In this case, the fuel is reprocessed to extract the fertile and the fissionable isotopes which then are transported to production places to obtain new fuel pellets. The materials remaining after reprocessing can be processed even more by recovering valuable isotopes for other uses.

The nuclear waste comprises many radioactive isotopes and transuranic elements which have an extremely long half-life time. Some of these elements are neptunium, plutonium, tectenium, and iodine. Keeping such elements for a long time implies vitrification or calcination, which embeds them in an amorphous matter with which they do not react for a long period of time.

An interesting aspect is that the radiation emitted by nuclear waste generates heat which is on the order of about 1.5 kW per ton of uranium in 10 years. This heat can in principle be used as a heat source either to generate power with low-temperature differential heat engines or for heating applications. However, there is a lack of confidence regarding the safety of using waste cooling systems as heat sources and such solutions have not been implemented to date.

After appropriate packing, the waste is deposited in long-term geological storage locations excavated in the form of tunnels in pits at more than 1000 m below the earth surface (see IAEA, 2011). Nuclear wastes can be categorized in three classes, namely:

• Low-level waste—which comprises the residual radioactive material from medical and industrial facilities which use radionuclides for various processes involving, for example, gamma- or X-ray generation. Some of the radioactive material disposed by other nuclear facilities is also low-level waste.

• High-level waste—generated during the nuclear fuel cycle, consisting of solid and liquid material which is highly radioactive.

• Transuranic waste—comprising radioactive materials with a decay time of longer than 20 years which emit alpha particles with sensible intensity.

Among all categories the safe disposal of high-level waste is the most important. These wastes are stored in underground repositories built in several places throughout the world, one of them being at Yucca Mountain in the United States, planned to accommodate 70 kt of nuclear waste. The countries that have national plans for nuclear waste repositories are the following:

• United States—storing the waste in tuff host rock in Yucca Mountain

• Finland—to be completed in 2020, storing the waste in granite host rock, at Olkiluoto location

• Germany—to be completed in 2020, with storage in salt, at Gorleben location

• France—to be completed by 2020, with storage in clay and granite

• Canada—to be completed in 2025, with storage in granite

• Japan—to be completed by 2030, with storage in sedimentary rock and granite

• Switzerland—with opening target in 2050, and storage in granite and clay

• Sweden—with completion target in 2020, and storage in granite host rock

There are various safety measures that must be taken to avoid direct exposure of humans and environment to radiation emitted by nuclear waste. Shielding is applied wherever possible. Also the spent fuel assemblies are placed in cooling pools and transferred into shielded protective canisters. The hazard that is possibly created by the nuclear waste is the dose that can be received through inhalation of escaped radionuclides. The maximum dose rate that a person can take is of 20–50 mSv per annum, where “mSv” stands for millisievert, which is the equivalent dose to a tissue, a calculation based on the absorbed radiation dose and factors dependent on the radiation type and other relevant parameters, such that the quantity represents a quantitative measure of radiation effects; note that 1 Sv = 1 m²s^− 2.

A brief comparison of nuclear fuel, fossil fuels, and biomass is given in Table 6.6. Definitively, the nuclear technology produces the most power per unit of hardware volume or mass. Coal and biomass combustion facilities are massive because they must include fuel handling facilities and large stacks; this decreases the energy density of the generated power. However, the Carnot factor, and therefore the exergy efficiency, associated with combustion technologies is the highest.

6.5 Nuclear-Based Cogeneration Systems

Cogeneration, which is the simultaneous production of thermal and electrical energy, has significant potential for transforming present energy supply systems so that they become more sustainable. In any type of thermal power plant a portion, usually 20–45%, of the input thermal energy is converted to electricity, and the remainder is rejected to the environment as waste heat. As nuclear power generation units can produce a high-pressure and high-temperature vapor, they can be utilized to generate heating, cooling, or freshwater with a topping cycle. In this chapter, we try to address the main nuclear-based cogeneration plants briefly. Besides power these cogeneration plants co-produce heating, freshwater, cooling, hydrogen, process heat, synthetic fuels or other commodities with market value. In Chapter 9 another perspective on the cogeneration system will be provided, when discussing multigeneration systems.

When cogeneration is applied better resource utilization is obtained. This is illustrated in Figure 6.25. In a cogeneration system the thermodynamic cycle can be connected to two heat sinks. One heat sink is the reference environment, typically air or a lake or river. The heat engine rejects heat at a temperature higher than the environment depending on the specific demand. In district heating or desalination applications the required temperature level is low (60–120 °C). In other applications heat can be required at intermediate or very high values. Obviously, the applications with a low-grade heat demand are the most favorable economically because valorizing the low grade heat will increase the return on investment in the nuclear power plant and make it more economically attractive.

A short numerical example can be given related to options (a) and (b) presented in Figure 6.26. Assume that there is a demand of 300 MW power and 350 MW heat. If heat and power are produced with separate technologies (for example, power is derived from a nuclear station and heat is generated by a natural-gas-burning furnace) then the following situation is valid: for 300 MW power the required thermal energy is 900 MW, and for 350 MW heating one needs to combust 410 MW equivalent natural gas for 85% furnace efficiency. Hence the overall efficiency is 650/1310 MW = 49.6%. With a cogeneration approach there is no need for natural gas, but rather 350 MW of heat is recovered from the nuclear power plant; note that in order to produce 300 MW power the power plant must reject into the environment about 600 MW of waste heat. Hence cogeneration is motivated by an efficiency of 650/900 MW = 72%.

Although there are some challenges, such as coping with the danger of radioactive contamination, there is obviously great potential for nuclear heat. Cogeneration with nuclear energy has been used in the past and is used today in a limited number of applications. One of the most common uses of nuclear-based cogeneration is for district heating. The worldwide experience with nuclear district heating measures 500 reactor-years, which represents only 4.1% of the global total of 12,193 reactor-years, according to IAEA (2006). Nuclear desalination has also been used in some countries with a cumulated experience of 250 reactor-years (~ 2% of global nuclear power station experience). Some limited experience exists for industrial heat cogeneration with nuclear reactors, namely for heavy water production, industrial steam cogeneration, salt refining, and cardboard production. Two main categories of heat demand can be satisfied through cogeneration:

• Tertiary sector, more specifically including heat demands in residential, commercial, governmental building settings (water and space heating or air conditioning).

• Primary and secondary sectors that require heat at a wide temperature range and capacities for water heating, space heating, drying processes, manufacturing, mining, agriculture, chemical processing, etc.

Figure 6.27 shows the temperature level with Generation IV nuclear reactors versus requirements of some relevant heat demanding processes. When heat is transported remotely from the nuclear reactor to the user site measures must be taken to avoid any possibility of radioactive contamination through the heat transfer fluid. Two configurations are suggested for reactor radioactive isolation, as shown in Figure 6.28. In the HLH configuration there is a higher pressure in the reactor loop and process loop. Possibly, radioactive contaminants may leak into the isolation loop; however, due to the pressure, contamination by radioactive particles remains confined in the isolation loop and does not pass further. The isolation loops may be treated for cleaning with specific methods. With the LHL configuration the isolation loop has the highest pressure and therefore will not allow penetration of particles or fluids from the other two loops.

Figure 6.29 suggests a possible linkage between a nuclear Rankine steam power plant and a heat- demanding system for cogeneration. The steam generator which produces superheated steam is part of the multistage Rankine power plant. A reheating circuit can be integrated with the steam generator. When cogeneration is required, a part of the superheated steam from the steam reheating circuit is directed toward a secondary low-pressure turbine (stream #27). Steam is expanded to an intermediate pressure in accordance with the temperature level required for cogeneration. After a heat transfer process with the cogeneration loop steam fully condenses and the liquid water is returned at the steam trap of the open heat water feeder. For applications where a higher temperature is required for the cogeneration loop the low-pressure turbine is not used, rather the cogeneration heat exchanger is supplied directly with high-temperature steam (Figure 6.29).

The power and heat products of a nuclear reactor station can be converted in industrial parks into multiple commodities such as steam, potable water, and hydrogen and used further for many processes, as suggested in Figure 6.30. Some examples of heat-demanding industrial processes are as follows:

• Aluminum production: the alumina-rich bauxite ore (Al₂O₃) must be extracted in slurry form mixed with sodium hydroxide at 110°–270 °C, with the intermediary products being sodium aluminate and aluminum hydroxide. Further, a rotating kiln is used for calcination of aluminum hydroxide at 1100 °C.

• Steam-coal gasification: this process can be conducted in two ways (i) steam-coal gasification and (ii) hydrogasification. In hydrogasification coal is combined with hydrogen in an exothermic reaction to produce synthetic natural gas. Nuclear heat is used to produce synthesis gas and extract hydrogen through steam reforming and water-gas shift reactions. The processes can be coupled with the VHTR for a required heat of 950 °C.

• Steam-assisted gravity drainage: this process is used to extract bituminous oil sands where steam at 10–15 MPa and 200–240 °C is required. In addition this process requires hydrogen for synthetic crude oil production.

As the population is growing, many regions are experiencing increased freshwater demands that greatly exceed the supply capability of existing infrastructures. The problem is compounded by increases in both the pollution and salinity of freshwater resources. Lack of freshwater is a prime factor inhibiting regional economic development. Seawater desalination is an important option for satisfying current and future demands for freshwater in arid regions in close proximity to the sea.

Desalination units need energy to separate salt from saline waters. The energy can be either heat for seawater distillation or mechanical energy to drive pumps for pressurization of seawater across membranes using reverse osmosis (RO). The major source of heat energy comes from fossil fuels, including coal, oil, and gas. Most of the large-size plants based on thermal and membrane processes are located near thermal power stations, which utilize fossil fuels to supply both steam and electrical power for desalination. However, the use of fossil fuel to produce freshwater has several side effects. As the combustion of fossil fuels emits greenhouse gases to the environment, it will lead to global warming, one of the main concerns of this century. In addition, these fuels are non-renewable and will be exhausted in next 50 years.

Therefore, because of their versatility as an energy source and hydrocarbon feedstock, there is a keen interest in conserving fossil fuels for other industrial applications, especially in countries with inadequate sources of energy. Over the long term it is not practical to establish a desalting economy based solely on fossil fuels because they have limited availability, whereas freshwater must continue to be available to sustain humankind. Long-term reliance on desalting should only occur if the availability of the energy supply is comparable to the required availability of the product. This criterion can be satisfied by nuclear energy. Interest in using nuclear energy for producing potable water has been growing worldwide in the past decade. This has been motivated by a wide variety of factors, from the economic competitiveness of nuclear energy to energy supply diversification; from conservation of limited fossil fuel resources to environmental protection; and by the spin-off effects of nuclear technology in industrial development. One of the main cogeneration units using nuclear power as a topping cycle is used to produce freshwater by multistageflash desalination techniques.

Hydrogen is expected to play a significant role as energy carrier in the future. Hydrogen can be used as fuel in almost every application where fossil fuels are utilized today. In contrast to fossil fuel, its combustion is without harmful emissions, disregarding NO_x emissions that can be effectively controlled. In addition, hydrogen can be transformed into useful forms of energy more efficiently than fossil fuels. Hydrogen is as safe as other common fuels, despite its common perception. Hydrogen is not an energy source and hence it does not exist in nature in its elemental form.

Therefore, hydrogen must be produced from water, the most abundant source of hydrogen, or from other sources. However, splitting of water for hydrogen production necessitates energy that is higher than the energy that can be obtained from the produced hydrogen. Therefore, hydrogen is considered an energy carrier for a suitable form of energy like electricity. It is commonly accepted that hydrogen is one of the promising energy carriers and that the demand for it will increase greatly in the near future, for it can be utilized as a clean fuel in diverse energy end-use sectors, including the conversion to electricity with no CO₂ emission. In addition, the nuclear fuel supply is estimated to be readily available even on a once-through fuel cycle for 50 to 100 years. Using breeder reactors and a closed fuel cycle, it is virtually inexhaustible. Because of its relative abundance, nuclear fuel is relatively cheap compared to fossil-based fuels.

Possible routes to generate hydrogen from water using nuclear energy are also depicted in Figure 6.31, with additional paths which assume processes other than water decomposition, such as coal gasification, natural gas reforming, petroleum naphtha reforming, or hydrogen sulfide cracking. In all such processes, high-temperature nuclear heat is used to conduct the required chemical reactions. Coal gasification and natural gas reforming methods include the water-gas shift reaction to generate additional hydrogen and reduce the carbon monoxide to CO₂. Nuclear-based reforming of fossil fuels is a promising method of high viability in the transition toward a fully implemented hydrogen economy which will reduce the world dependence on fossil fuel energy.

Although fossil fuel reforming can be used to generate hydrogen, it is attractive to use nuclear heat to synthesize Fischer–Tropsch fuels or methanol. Such fuels can also be used for transportation vehicles. With this method, nuclear heat is used to generate synthesis gas, and then to generate either diesel or methanol. Another process of great interest is fertilizer production, such as ammonia and urea, with nuclear energy. A nuclear ammonia/urea production facility would comprise the nuclear reactor, a heat transfer system for high-temperature nuclear process heat, a nuclear power plant, a nuclear hydrogen generation unit, an air separation unit, H₂/CO₂ separators, a Haber–Bosch synthesis unit, and an NH₃/CO₂ reactor to generate urea.

Hydrogen sulfide is a potential source of hydrogen. Hydrogen sulfide can be extracted from some geothermal wells, oil wells, volcano sites, and seas. The Black Sea represents the highest reserve of hydrogen sulfide in the world. In the Black Sea, the hydrogen sulfide can be viewed as a renewable resource because it is generated by microorganisms under the influence of solar radiation. Typically, thermo-catalytic cracking of H₂S can be applied to generate hydrogen and sulfur, as two valuable products using nuclear energy.

Case Study 6.3: Integrated Desalination and Hydrogen Production with Nuclear Reactor

In this example, which is based on the work by Orhan et al. (2010), we analyze a coupling of the Cu–Cl cycle with a desalination plant for hydrogen production from nuclear energy and seawater. To produce hydrogen by splitting the water molecule one needs to supply freshwater to the hydrogen production facility. Thus, to avoid causing one problem while solving another problem, hydrogen could be produced from seawater rather than limited freshwater sources. Nuclear energy is used to drive the process. Here we consider five configurations, in hopes of determining or helping to determine an optimum option to couple the Cu–Cl cycle with a desalination plant.

Configuration 1: the Cu–Cl cycle is coupled to a desalination plant using nuclear energy, as shown in Figure 6.32. Salty water is put into the desalination plant and salt is removed from the water using waste energy from the nuclear reactor moderator at 70–80 °C. Freshwater supplied by the desalination plant is decomposed into hydrogen and oxygen by the Cu–Cl cycle driven by nuclear energy. humidification–dehumidification (HD) technology is used for desalination.

Configuration 2: the thermal energy recovered from the Cu–Cl cycle is transferred to the desalination plant to remove salt from freshwater. As illustrated in Figure 6.33, the desalination plant operates as a subsystem of the Cu–Cl cycle. Considering the overall system, only process/waste energy from the nuclear reactor and salty water enter the system. Hydrogen is produced, and oxygen and salt are byproducts. A drawback to this configuration is the efficiency decrease (~ 3–5%) incurred by the Cu–Cl cycle as the recovered energy is used for the desalination process rather than within the cycle itself. But the Cu–Cu cycle is a subsection of the plant and, when the desalination plant and the Cu–Cl cycle are considered as a combined system, the efficiency is not affected significantly as the recovered energy is used within the overall system. The efficiency of the combined system (desalination plant and copper chlorine cycle) is about 0.4. Multiple effect desalination technology is used.

Configuration 3: the nuclear thermal energy is used directly in the desalination plant and to drive the hydrogen plant as indicated in Figure 6.34. The thermal energy recovered in the Cu–Cl cycle is used within that cycle. A desalination method with a high capacity and low production cost is used as high-grade energy is available. Multistage flash desalination method is used.

Configuration 4: in addition to nuclear thermal energy, solar energy is used to complete the energy requirement for the process. The solar energy drives directly and completely the desalination plant which supplies the hydrogen production plant with freshwater. Process and waste energy from a nuclear plant are used in the Cu–Cl cycle as indicated in Figure 6.35. Vapor compression (VC) desalination technology is applied.

A drawback of this configuration is its dependence on the availability of solar energy, which is intermittent. As the capacity of the desalination plant and hence the Cu–Cl cycle determines the required capacity of the solar collectors, the site for the Cu–Cl cycle and desalination plant is chosen carefully. However, as constraints likely exist regarding the location of the nuclear reactor, this configuration is likely advantageous if the nuclear reactor is located in an area with high solar insolation. Otherwise, a desalination process with low capacity and low inlet energy requirements is used. RO technology is applied.

Configuration 5: off-peak electricity from a nuclear reactor is used to operate desalination plant. When off-peak electricity is available, it is usually less expensive than peak electricity and thus is used by many industrial processes. Off-peak electricity can also be used to desalinate seawater and to produce hydrogen, and could be beneficial for the same reason. This configuration includes a Cu–Cl cycle coupled with a desalination plant driven by off-peak electricity (as in Figure 6.36). In this case, efficient membrane processes are used as electrical energy is supplied.

The overall energy efficiency of the coupled system, η_overall, represents the fraction of energy supplied to produce hydrogen from salted water that is recovered in the energy content of H₂ based on its lower heating value. The total energy required to produce hydrogen from salty water can be written as

$Q_{\mathrm{in},\mathrm{total}}=Q_{\mathrm{in},\mathrm{Cu}-\mathrm{Cl}}+Q_{\mathrm{in},\mathrm{Desalination}}$

  (6.12)

and the overall efficiency is

${\eta }_{\mathrm{overall}}=\mathrm{LHV}/Q_{\mathrm{in},\mathrm{total}},$

  (6.13)

where LHV is the lower heating value of hydrogen.

The variation of costs for hydrogen production capacity at a Cu–Cl thermochemical plant is shown in Figure 6.36. These costs are based on producing hydrogen from freshwater and do not include costs associated with desalination. As can be observed, the capital cost of the Cu–Cl cycle varies from 1.8 to 0.3 $/kg H_2, based on the capacity of the cycle. The capital cost of the cycle per unit of hydrogen output is less for a larger capacity plant while the production cost (energy cost) remains constant at around 0.9 $/kg H₂, mainly because the reaction energy (per unit hydrogen produced) of any chemical or physical reaction occurring in the Cu–Cl cycle does not change based on plant capacity. The total cost of hydrogen production is the sum of the capital cost, energy cost, and also some additional costs such as operating, storage, and distributions costs. In Figure 6.37, two main cost items (capital and energy costs) are given, along with the total cost to build a pilot plant using the Cu–Cl cycle, presumably at the commercialization stage. The total cost varies from 3.5 to 2 $/kg H_2, in an inversely proportional relationship with plant capacity.

The total cost of the desalination plant is shown in Figure 6.38. The least expensive desalination method is HD at 7.5 $/(kg H₂/day), which is considered in Configuration 1, and the technology with the highest initial cost (11.9 $/(kg H₂/day)) is multistage flash desalination, which is used in Configuration 3. The total costs of VC and RO appear to be the same at 9.62 $/(kg H₂/day), while the capital cost of multiple effect desalination is 11.7 $/(kg H₂/day).

The energy efficiencies of the Cu–Cl cycle, desalination plant, and the overall system, including the Cu–Cl cycle and the desalination plant, are shown in Figure 6.39. The effect of the Cu–Cl cycle on the overall system is dominant, as the desalination plant uses much less energy than the Cu–Cl cycle. Thus, the efficiency of the Cu–Cl cycle and the overall system are very similar for each case. Configuration 1 exhibits a higher efficiency, as waste heat from the nuclear reactor (which is assumed to be “free”) is utilized in this option. In Configuration 2, recovered heat from the Cu–Cl cycle is used for desalination instead of internally, within the cycle. Thus, the Cu–Cl cycle and the overall system operate at lower efficiencies, as the effect of the cycle is very important on the overall system.