Electrical-energy consumption per capita in selected countries (Wikipedia, 2012).
It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living (see Table 1). In general, electrical energy can be produced by: 1) non-renewable sources such as coal, natural gas, oil, and nuclear; and 2) renewable sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy production are: 1) thermal - primary coal and secondary natural gas; 2) nuclear and 3) hydro. The rest of the sources might have visible impact just in some countries (see Figure 1). In addition, the renewable sources such as wind (see Figure 1b,c) and solar are not really reliable sources for industrial power generation, because they depend on Mother nature and relative costs of electrical energy generated by these and some other renewable sources with exception of large hydro-electric power plants can be significantly higher than those generated by non-renewable sources. Therefore, thermal and nuclear electrical-energy production will be considered further.
2. Thermal power plants
In general, the major driving force for all advances in thermal and Nuclear Power Plants (NPPs) is thermal efficiency. Ranges of thermal efficiencies of modern power plants are listed in Table 2 for references purposes.
|1||Combined-cycle power plant (combination of Brayton gas-turbine cycle (fuel natural or Liquefied Natural Gas (LNG); combustion-products parameters at the gas-turbine inlet:
||Up to 62|
|2||Supercritical-pressure coal-fired thermal power plant (new plants) (Rankine-cycle steam inlet turbine parameters:
||Up to 55|
|3||Subcritical-pressure coal-fired thermal power plant (older plants) (Rankine-cycle steam:
||Up to 40|
|4||Carbon-dioxide-cooled reactor (Advanced Gas-cooled Reactor (AGR) (see Figure 12)) NPP (Generation III, current fleet) (reactor coolant – carbon dioxide:
||Up to 42|
|5||Sodium-cooled Fast Reactor (SFR) NPP (see Figure 15) (Generation III and IV, currently just one reactor – BN-600 operates in Russia) (reactor coolant – liquid sodium:
||Up to 40|
|6||Pressurized Water Reactor (PWR) NPP (Generation III+, to be implemented within next 1–10 years) (reactor coolant – light water:
||Up to 36-38|
|7||PWR NPP (see Figure 9) (Generation III, current fleet) (reactor coolant – light water:
|8||Boiling Water Reactor (BWR) NPP (see Figure 10) (Generation III, current fleet) (reactor coolant light water; direct cycle; steam parameters at the turbine inlet:
|9||RBMK reactor (boiling reactor, pressure-channel design) NPP (see Figure 14) (Generation II and III, current fleet) (reactor coolant light water; direct cycle; steam parameters at the turbine inlet:
|10||Pressurized Heavy Water Reactor (PHWR) NPP (see Figure 11) (Generation III, current fleet) (reactor coolant – heavy water:
3. Coal-fired thermal power plants
For thousands years, mankind used and still is using wood and coal for heating purposes. For about 100 years, coal is used for generating electrical energy at coal-fired thermal power plants worldwide. All coal-fired power plants (see Figure 2) operate based on, so-called, steam Rankine cycle, which can be organized at two different levels of pressures: 1) older or smaller capacity power plants operate at steam pressures no higher than 16 – 17 MPa and 2) modern large capacity power plants operate at supercritical pressures from 23.5 MPa and up to 38 MPa (see Figure 3). Supercritical pressures See some explanations on supercritical-pressures specifics at the end of this section.
See some explanations on supercritical-pressures specifics at the end of this section.
In spite of advances in coal-fired power-plants design and operation worldwide they are still considered as not environmental friendly due to producing a lot of carbon-dioxide emissions as a result of combustion process plus ash, slag and even acid rains (Pioro et al., 2010). However, it should be admitted that known resources of coal worldwide are the largest compared to that of other fossil fuels (natural gas and oil).
For better understanding specifics of supercritical water compared to water at subcritical pressures it is important to define special terms and expressions used at these conditions. For better understanding of these terms and expressions Figures 4 – 7 are shown below.
4. Definitions of selected terms and expressions related to critical and supercritical regions (Pioro and Mokry, 2011a)
General trends of various properties near the critical and pseudocritical points (Pioro et al., 2011; Pioro and Mokry, 2011a; Pioro and Duffey, 2007) can be illustrated on a basis of those of water. Figure 5 shows variations in basic thermophysical properties of water at a supercritical pressure of 25 MPa (also, in addition, see Figure 6). Thermophysical properties of 105 pure fluids including water, carbon dioxide, helium, refrigerants, etc., 5 pseudo-pure fluids (such as air) and mixtures with up to 20 components at different pressures and temperatures, including critical and supercritical regions, can be calculated using the NIST REFPROP software (2010).
At critical and supercritical pressures a fluid is considered as a single-phase substance in spite of the fact that all thermophysical properties undergo significant changes within critical and pseudocritical regions (see Figure 5). Near the critical point, these changes are dramatic. In the vicinity of pseudocritical points, with an increase in pressure, these changes become less pronounced (see Figure 6).
At supercritical pressures properties such as density (see Figure 5) and dynamic viscosity undergo a significant drop (near the critical point this drop is almost vertical) within a very narrow temperature range, while the kinematic viscosity and specific enthalpy (see Figure 5) undergo a sharp increase. The volume expansivity, specific heat, thermal conductivity and Prandtl number have peaks near the critical and pseudocritical points (see Figures 5 and 6). Magnitudes of these peaks decrease very quickly with an increase in pressure (see Figure 6). Also, “peaks” transform into “humps” profiles at pressures beyond the critical pressure. It should be noted that the dynamic viscosity, kinematic viscosity and thermal conductivity (see Figure 5) undergo through the minimum right after critical and pseudocritical points.
The specific heat of water (as well as of other fluids) has a maximum value in the critical point. The exact temperature that corresponds to the specific-heat peak above the critical pressure is known as a pseudocritical temperature (see Figure 4). At pressures approximately above 300 MPa (see Figure 6) a peak (here it is better to say “a hump”) in specific heat almost disappears, therefore, such term as a
In general, crossing the pseudocritical line from left to right (see Figure 4) is quite similar as crossing the saturation line from liquid into vapour. The major difference in crossing these two lines is that all changes (even drastic variations) in thermophysical properties at supercritical pressures are gradual and continuous, which take place within a certain temperature range (see Figure 5). On the contrary, at subcritical pressures there is properties discontinuation on the saturation line: one value for liquid and another for vapour (see Figure 7). Therefore, supercritical fluids behave as single-phase substances (Gupta et al., 2012). Also, when dealing with supercritical fluids we usually apply the term “
5. Combined-cycle thermal power plants
Natural gas is considered as a relatively “clean” fossil fuel compared to coal and oil, but still emits a lot of carbon dioxide due to combustion process when it used for electrical generation. The most efficient modern thermal power plants with thermal efficiencies within a range of 50 – 62% are, so-called, combined-cycle power plants, which use natural gas as a fuel (see Figure 8).
In spite of advances in thermal power plants design and operation, they still emit carbon dioxide into atmosphere, which is currently considered as one of the major reasons for a climate change. In addition, all fossil-fuel resources are depleting quite fast. Therefore, a new reliable and environmental friendly source for the electrical-energy generation should be considered.
6. Nuclear power plants
6.1. Modern nuclear reactors
Nuclear power is also a non-renewable source as the fossil fuels, but nuclear resources can be used for significantly longer time than some fossil fuels plus nuclear power does not emit carbon dioxide into atmosphere. Currently, this source of energy is considered as the most viable one for electrical generation for the next 50 – 100 years.
For better understanding specifics of current and future nuclear-power reactors it is important to define their various classifications.
6.2. Classifications of nuclear-power reactors
By neutron spectrum: (a) thermal (the vast majority of current nuclear-power reactors), (b) fast (currently, only one nuclear-power reactor is in operation in Russia: SFR – BN-600), and (c) interim or mixed spectrum.
By reactor-core design:
Neutron-core design: (a) homogeneous, i.e., the fuel and reactor coolant are mixed together (one of the Generation IV nuclear-reactors concepts) and (b) heterogeneous, i.e., the fuel and reactor coolant are separated through a sheath or cladding (currently, all nuclear-power reactors);
General core design: (a) Pressure-Vessel (PV) (the majority of current nuclear-power reactors including PWRs, BWRs, etc.) and (b) Pressure-Channel (PCh) or Pressure-Tube (PT) reactors (CANDU ((CANada Deuterium-Uranium) reactors, RBMKs), EGPs (Power Heterogeneous Loop reactor (in Russian abbreviations)), etc.).
Water-cooled reactors: (a) Light-Water (H2O) Reactors (LWRs) - PWRs, BWRs, RBMKs, EGPs, and (b) heavy-water (D2O) reactors – mainly CANDU-type reactors.
Gas-cooled reactors: Carbon-dioxide-cooled reactors (Magnox
reactors (Gas-Cooled Reactors (GCRs)) and AGRs) and helium-cooled reactors (two Generation IV nuclear-reactor concepts); (c) liquid-metal-cooled reactors: SFR, lead-cooled and lead-bismuth-cooled reactors (Generation IV nuclear-reactor concepts); (d) molten-salt-cooled reactors (one of Generation IV nuclear-reactor concepts); and (e) organic-fluids-cooled reactors (existed only as experimental reactors some time ago).
In this reactor the fuel-rod sheath is made of magnesium alloy known by the trade name as “Magnox”, which was used as the name of the reactor (Hewitt and Collier, 2000).
By type of a moderator (Kirillov et al., 2007): (a) liquid moderator (H2O and D2O are currently used in nuclear-power reactors as moderators) and (b) solid moderator (graphite
(RBMKs, EGPs, Magnox reactors (GCRs), AGRs), zirconium hydride (ZrH2), beryllium (Be) and beryllium oxide (BeO)).
After the Chernobyl NPP severe nuclear accident in Ukraine in 1986 with the RBMK reactor, graphite is no longer considered as a possible moderator in any water-cooled reactors.
By application: (a) power reactors (PWRs, BWRs, CANDU reactors, GCRs, AGRs, RBMKs, EGPs, SFR from current fleet) (b) research reactors (for example, NRU (National Research Universal) (AECL, Canada, http://www.aecl.ca/Programs/NRU.htm), etc.), (c) transport or mobile reactors (submarines and ships (icebreakers, air-carriers, etc.), (d) industrial reactors for isotope production (for example, NRU), etc., and (e) multipurpose reactors (for example, NRU, etc.).
By number of flow circuits: (a) single-flow circuit (once-through or direct-cycle reactors) (BWRs, RBMKs, EGPs); (b) double-flow circuit (PWRs, PHWRs, GCRs, AGRs) and (c) triple-flow circuit (usually SFRs).
By fuel enrichment: (a) Natural-Uranium fuel (NU) (99.3%wt of non-fissile isotope uranium-238 (238U) and 0.7% of fissile isotope uranium-235 (235U)) (CANDU-type reactors, Magnox reactors), (b) Slightly-Enriched Uranium (SEU) (0.8 – 2%wt of 235U), (c) Low-Enriched Uranium (LEU) (2 – 20% of 235U) (the vast majority of current nuclear-power reactors: PWRs, BWRs, AGRs, RBMKs, EGPs), and (d) Highly-Enriched Uranium (HEU) (>20%wt of 235U) (can be SFR).
By used fuel (Peiman et al., 2012): (a) Conventional nuclear fuels (low thermal conductivity): Uranium dioxide (UO2, used in the vast majority of nuclear-power reactors), Mixed OXides (MOX) ((U0.8Pu0.2)O2, where 0.8 and 0.2 are the molar parts of UO2 and PuO2, used in some reactors) and thoria (ThO2) (considered for a possible use instead of UO2 in some countries, usually, with large resources of this type of fuel, for example, in India); and (b) alternative nuclear fuels (high thermal conductivity): Uranium dioxide plus silicon carbide (UO2–SiC), uranium dioxide composed of graphite fibre (UO2–C), uranium dioxide plus beryllium oxide (UO2–BeO), uranium dicarbide (UC2), uranium monocarbide (UC) and uranium mononitride (UN); the last three fuels are mainly intended for use in high-temperature Generation IV reactors.
First success of using nuclear power for electrical generation was achieved in several countries within 50-s, and currently, Generations II and III nuclear-power reactors are operating around the world (see Tables 3 and 4 and Figures 9-15). In general, definitions of nuclear-reactors generations are as the following: 1) Generation I (1950 – 1965) – early prototypes of nuclear reactors; 2) Generation II (1965 – 1995) – commercial power reactors; 3) Generation III (1995 – 2010) – modern reactors (water-cooled NPPs with thermal efficiencies within 30 – 36%; carbon-dioxide-cooled NPPs with the thermal efficiency up to 42% and liquid sodium-cooled NPPs with the thermal efficiency up to 40%) and Generation III+ (2010 – 2025) – reactors with improved parameters (evolutionary design improvements) (water-cooled NPPs with the thermal efficiency up to 38%) (see Table 5); and 4) Generation IV (2025 - …) – reactors in principle with new parameters (NPPs with the thermal efficiency of 43 – 50% and even higher for all types of reactors).
|1. PWRs (see Figure 9 and Tables 6 and 7) – 267 (
|2. BWRs or ABWRs (see Figure 10 and Table 8) – 84 (
|3. GCRs (see Figures 12 and 13) – 17 (
|4. PHWRs (see Figure 11) – 51 (
|5. Light-water, Graphite-moderated Reactors (LGRs) (see Figure 14 and Table 6) – 15 (10 GWel), Russia, 11 RBMKs and 4 EGPs1 (earlier prototype of RBMK).|
|6. Liquid-Metal Fast-Breeder Reactors (LMFBRs) (see Figure 15 and Table 6) – 1 (0.6 GWel), SFR, Russia; forthcoming – 4 (1.5 GWel).|
|5.||S. Korea||21 (
|10.||China 14 (
|ABWR – Toshiba, Mitsubishi Heavy Industries and Hitachi-GE (Japan-USA) (the only one Generation III+ reactor design already implemented in the power industry).|
|Advanced CANDU Reactor (ACR-1000) AECL, Canada.|
|Advanced Plant (AP-1000) – Toshiba-Westinghouse (Japan-USA) (6 under construction in China and 6 planned to be built in China and 6 – in USA).|
|Advanced PWR (APR-1400) – South Korea (4 under construction in S. Korea and 4 planned to be built in United Arad Emirates).|
|European Pressurized-water Reactor (EPR) AREVA, France (1 should be put into operation in Finland, 1 under construction in France and 2 in China and 2 planned to be built in USA).|
|VVER1 (design AES2-2006 or VVER-1200 with ~1200 MWel) – GIDROPRESS, Russia (2 under construction in Russia and several more planned to be built in various countries). Reference parameters of Generation III+ VVER (Ryzhov et al., 2010) are listed below:|
|Thermal power, MWth||3200|
|Electric power, MWel||1160|
|NPP thermal efficiency, %||36|
|Primary coolant pressure, MPa||16.2|
|Steam-generator pressure, MPa||7.0|
|Coolant temperature at reactor inlet, oC||298|
|Coolant temperature at reactor outlet, oC||329|
|NPP service life, years||50|
|Main equipment service life, years||60|
|Replaced equipment service life, years, not less than||30|
|Capacity factor, %||up to 90|
|Load factor, %||up to 92|
|Equipment availability factor||99|
|Length of fuel cycle, years||4-5|
|Frequency of refuellings, months||12-18|
|Fuel assembly maximum burn-up, MW day/kgU||up to 60-70|
|Inter-repair period length, years||4-8|
|Annual average length of scheduled shut-downs (for refuellings, scheduled maintenance work), days per year||16-40|
|Refueling length, days per year||≤16|
|Number of not scheduled reactor shutdowns per year||≤1|
|Frequency of severe core damage, 1/year||<106|
|Frequency of limiting emergency release, 1/year||<107|
|Efficient time of passive safety and emergency control system operation without operator’s action and power supply, hour||≥24|
|OBE/SSE, magnitude of MSK-64 scale||6 and 7*|
|Compliance with EUR requirements, yes/no||Yes|
|*RP main stationary equipment is designed for SSE of magnitude 8.|
|Thermal power, MWth||1375||3000||62||3200||1500|
|Electrical power, MWel||440||1000||12||1000||600|
|Thermal efficiency, %||32.0||33.3||19.3||31.3||40.0|
|Coolant pressure, MPa||12.3||15.7||6.2||6.9||~0.1|
|Coolant massflow rate, t/s||11.3||23.6||0.17||13.3||6.9|
|Coolant inlet/outlet temperatures, °C||270/298||290/322||265||284||380/550|
|Steam massflow rate, t/s||0.75||1.6||0.026||1.56||0.18|
|Steam pressure, MPa||4.3||5.9||6.5||6.6||15.3|
|Steam temperature, °C||256||276||280||280||505|
|Reactor core: Diameter/Height m/m||3.8/11.8||4.5/10.9||4.2/3.0||11.8/7||2.1/0.75|
|Fuel enrichment, %||3.6||4.3||3.0;3.6||2.0-2.4||21;29.4|
|No. of fuel bundles||349||163||273||1580||369|
|Pressure Vessel (PV) ID, m||3.91|
|PV wall thickness, m||0.19|
|PV height without cover, m||10.8|
|Core equivalent diameter, m||2.88|
|Core height, m||2.5|
|Volume heat flux, MW/m3||83|
|No. of fuel assemblies||349|
|No of rods per assembly||127|
|Fuel mass, ton||42|
|Part of fuel reloaded during year||1/3|
|Thermal output, MWth||3830|
|Electrical output, MWe||1330|
|Thermal efficiency, %||34|
|Specific power, kW/kg(U)||26|
|Power density, kW/L||56|
|Average linear heat flux, kW/m||20.7|
|Fuel-rod heat flux average/max, MW/m2||0.51/1.12|
|Core massflow rate, kg/s||14,167|
|Core void fraction average/max||0.37 / 0.75|
|Feedwater inlet temperature, °C||216|
|Steam outlet temperature, °C||290|
|Steam outlet massflow rate, kg/s||2083|
|Inside Diameter, m||6.4|
|Wall thickness, m||0.15|
|Pellet OD, mm||10.6|
|Fuel rod OD, mm||12.5|
|Zircaloy sheath (cladding) thickness, mm||0.86|
Analysis of data listed in Table 3 shows that the vast majority nuclear reactors are water-cooled units. Only reactors built in UK are the gas-cooled type, and one reactor in Russia uses liquid sodium for its cooling.
UK carbon-dioxide-cooled reactors consist of two designs (Hewitt and Collier, 2000): 1) older design – Magnox reactor (GCR) (see Figure 13) and 2) newer design – AGR (see Figure 12). The Magnox design is a natural-uranium graphite-moderated reactor with the following parameters: Coolant – carbon dioxide; pressure - 2 MPa; outlet/inlet temperatures – 414/250°C; core diameter – about 14 m; height – about 8 m; magnesium-alloy sheath with fins; and thermal efficiency – about 32%. AGRs have the following parameters: Coolant – carbon dioxide; pressure - 4 MPa; outlet/inlet temperatures – 650/292°C; secondary-loop steam – 17 MPa and 560°C; stainless-steel sheath with ribs and hollow fuel pellets (see Figure 16b); enriched fuel 2.3%; and thermal efficiency – about 42% (the highest in nuclear-power industry so far). However, both these reactor designs will not be constructed anymore. They will just operate to the end of their life term and will be shut down. The same is applied to Russian RBMKs and EGPs.
Just for reference purposes, typical fuel elements (rods) / bundles of various reactors are shown in Figure 16, and typical sheath temperatures, heat transfer coefficients and heat fluxes are listed below.
Typical maximum sheath temperatures for steady operation (Hewitt and Collier, 2000)
All current NPPs and oncoming Generation III+ NPPs are not very competitive with modern thermal power plants in terms of their thermal efficiency, a difference in values of thermal efficiencies between thermal and NPPs can be up to 20 – 30% (see Table 2). Therefore, new generation NPPs should be designed and built in the nearest future.
7. Next generation nuclear reactors
The demand for clean, non-fossil based electricity is growing; therefore, the world needs to develop new nuclear reactors with higher thermal efficiencies in order to increase electricity generation and decrease detrimental effects on the environment. The current fleet of NPPs is classified as Generation II and III (just a limited number of Generation III+ reactors (mainly, ABWRs) operates in some countries). However, all these designs (here we are talking about only water-cooled power reactors) are not as energy efficient as they should be, because their operating temperatures are relatively low, i.e., below 350°C for a reactor coolant and even lower for steam.
Currently, a group of countries, including Canada, EU, Japan, Russia, USA and others have initiated an international collaboration to develop the next generation nuclear reactors (Generation IV reactors). The ultimate goal of developing such reactors is an increase in thermal efficiencies of NPPs from 30 – 36% to 45 - 50% and even higher. This increase in thermal efficiency would result in a higher production of electricity compared to current LWR technologies per 1 kg of uranium.
The Generation IV International Forum (GIF) Program has narrowed design options of nuclear reactors to six concepts. These concepts are: 1) Gas-cooled Fast Reactor (GFR) or just High Temperature Reactor (HTR), 2) Very High Temperature Reactor (VHTR), 3) Sodium-cooled Fast Reactor (SFR), 4) Lead-cooled Fast Reactor (LFR), 5) Molten Salt Reactor (MSR), and 6) SuperCritical Water-cooled Reactor (SCWR). Figures 17 – 24 show schematics of these concepts. These nuclear-reactor concepts differ one from each other in terms of their design, neutron spectrum, coolant, moderator, operating temperatures and pressures. A brief description of each Generation IV nuclear-reactor concept has been provided below.
|Coolant inlet/outlet temperatures||°C||490/850|
|Coolant massflow rate||kg/s||320|
|Average power density||MWth/m3||100|
|Reference fuel compound||UPuC/SiC (70/30%) with about 20% Pu|
Gas-cooled Fast Reactor (GFR) or High Temperature Reactor (HTR) (see Figure 17 and Table 9.) is a fast-neutron-spectrum reactor, which can be used for the production of electricity and co-generation of hydrogen through thermochemical cycles or high-temperature electrolysis. The coolant is helium with inlet and outlet temperatures of 490 and 850°C, respectively. The net plant efficiency is about 48% with the direct Brayton helium-gas-turbine cycle. Table 9 lists a summary of design parameters for GFR (US DOE, 2002). However, due to some problems with implementation of the direct Brayton helium-gas-turbine cycle, the indirect Rankine steam cycle or even indirect supercritical carbon-dioxide Brayton gas-turbine cycle are also considered. The indirect cycles will be linked to the GFR through heat exchangers.
Very High Temperature Reactor (VHTR) (see Figure 18) is a thermal-neutron-spectrum reactor. The ultimate purpose of this nuclear-reactor design is the co-generation of hydrogen through high-temperature electrolysis. In a VHTR, graphite and helium have been chosen as the moderator and the coolant, respectively. The inlet and outlet temperatures of the coolant are 640 and 1000°C, respectively, at a pressure of 7 MPa (US DOE, 2002). Due to such high outlet temperatures, the thermal efficiency of VHTR will be above 50%. A summary of design parameters of VHTR are listed in Table 10 (US DOE, 2002).
In general, the US DOE supports research on several Generation IV reactor concepts (http://nuclear.energy.gov/genIV/neGenIV4.html). However, the priority is being given to the VHTR, as a system compatible with advanced electricity production, hydrogen co-generation and high-temperature process-heat applications.
|Average power density||MWth/m3||610|
|Coolant inlet/outlet temperatures||°C||640/1000|
|Reference fuel compound||ZrC-coated particles in pins or pebbles|
Similar to GFR, SFR (see Figure 19) is a fast-neutron-spectrum reactor. The main objectives of SFR are the management of high-level radioactive wastes and production of electricity. SFR uses liquid sodium as a reactor coolant with an outlet temperature between 530 and 550°C at the atmospheric pressure. The primary choices of fuel for SFR are oxide and metallic fuels. Table 11 lists a summary of design parameters of SFR (US DOE, 2002). The SFR concept is also on the priority list for the US DOE (http://nuclear.energy.gov/genIV/neGenIV4.html).
Currently, SFR is the only one Generation IV concept implemented in the power industry. Russia and Japan are leaders within this area. In particularly, Russia operates SFR at the Beloyarsk NPP (for details, see BN-600 in Table 6) and constructs even more powerful SFR – BN-850. Japan has operated SFR at the Monju NPP some time ago (http://en.wikipedia.org/wiki/Monju_Nuclear_Power_Plant). In Russia and Japan the SFRs are connected to the subcritical-pressure Rankine steam cycle through heat exchangers (see Figure 19). However, in the US and some other countries a supercritical carbon-dioxide Brayton gas-turbine cycle is considered as the power cycle for future SFRs, because carbon dioxide and sodium are considered to be more compatible than water and sodium. In general, sodium is highly reactive metal. It reacts with water evolving hydrogen gas and releasing heat. Due to that sodium can ignite spontaneously with water. Also, it can ignite and burn in air at high temperatures. Therefore, special precautions should be taken for safe operation of this type reactor. One of them is the triple-flow circuit with the intermediate sodium loop between the reactor coolant (primary sodium) and water as the working fluid in the power cycle.
|Coolant melting/boiling temperatures||°C||98/883|
|Coolant density at 450°C||kg/m3||844|
|Pressure inside reactor||MPa||~0.1|
|Coolant maximum outlet temperature||°C||530–550|
|Average power density||MWth/m3||350|
|Reference fuel compound||Oxide or metal alloy|
|Cladding||Ferritic or ODS ferritic|
|Reactor power (thermal/electrical)||MW||700/300||2800/1200|
|Coolant melting/boiling temperatures||°C||328/1743|
|Coolant density at 450°C||kg/m3||10,520|
|Pressure inside reactor||MPa||~0.1|
|Coolant inlet/outlet temperatures||°C||420/540|
|Coolant massflow rate||t/s||40||158|
|Maximum coolant velocity||m/s||1.8||1.7|
|Term of fuel inside reactor||years||5||5–6|
|Fuel reloading per year||1|
|Core diameter/height||m / m||2.3/1.1||4.8/1.1|
|Number of fuel bundles||185||332|
|Fuel-rod diameter||mm||9.1; 9.6; 10.4|
|Maximum cladding temperature||°C||650|
|Steam-generator inlet/outlet temperatures||°C||340/520|
|Term of reactor||years||30||60|
LFR (see Figure 20) is a fast-neutron-spectrum reactor, which uses lead or lead-bismuth as the reactor coolant. The outlet temperature of the coolant is about 550°C (but can be as high as 800°C) at an atmospheric pressure. The primary choice of fuel is a nitride fuel. The supercritical carbon-dioxide Brayton gas-turbine cycle has been chosen as a primary choice for the power cycle in US and some other countries, while the supercritical-steam Rankine cycle is considered as the primary choice in Russia (see Table 12).
MSR (see Figure 21) is a thermal-neutron-spectrum reactor, which uses a molten fluoride salt with dissolved uranium while the moderator is made of graphite. The inlet temperature of the coolant (e.g., fuel-salt mixture) is 565°C while the outlet temperature reaches 700°C. However, the outlet temperature of the fuel-salt mixture can even increase to 850°C when co-generation of hydrogen is considered as an option. The thermal efficiency of the plant is between 45 and 50%. Table 13 lists the design parameters of MSR (US DOE, 2002).
|Net thermal efficiency||%||4450|
|Average power density||MWth/m3||22|
|Fuel-salt inlet/outlet temperatures||°C||565/700 (800)|
The design of SCWRs is seen as the natural and ultimate evolution of today’s conventional water-cooled nuclear reactors (Schulenberg and Starflinger, 2012; Pioro, 2011; Oka et al., 2010; Pioro and Duffey, 2007):
Modern PWRs operate at pressures of 15 − 16 MPa.
BWRs are the once-through or direct-cycle design, i.e., steam from a nuclear reactor is forwarded directly into a turbine.
Some experimental reactors used nuclear steam reheat with outlet steam temperatures well beyond the critical temperature, but at pressures below the critical pressure (Saltanov and Pioro, 2011). And
Modern supercritical-pressure turbines, at pressures of about 25 MPa and inlet temperatures of about 600°C, operate successfully at coal-fired thermal power plants for more than 50 years.
|Coolant inlet/outlet temperatures||ºC||280/550||280/530||280/500|
|Maximum cladding temperature||ºC||630||630||−|
In general, SCWRs can be classified based on a pressure boundary, neutron spectrum and/or moderator (Pioro and Duffey, 2007). In terms of the pressure boundary, SCWRs are classified into two categories, a) Pressure Vessel (PV) SCWRs (see Figure 22), and b) Pressure Tube (PT) or Pressure Channel (PCh) SCWRs (see Figures 23 and 24). The PV SCWR requires a pressure vessel with a wall thickness of about 50 cm in order to withstand supercritical pressures. Figure 22 shows a scheme of a PV SCWR NPP. Table 14 lists general operating parameters of modern PV-SCWR concepts. On the other hand, the core of a PT SCWR consists of distributed pressure channels, with a thickness of about 10 mm, which might be oriented vertically or horizontally, analogous to CANDU and RBMK reactors, respectively. For instance, SCW CANDU reactor (Figure 23) consists of 300 horizontal fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at a pressure of 25 MPa (Pioro and Duffey, 2007). It should be noted that a vertical-core option (Figure 24) has not been ruled out; both horizontal and vertical cores are being studied by the Atomic Energy of Canada Limited (AECL). Table 15 provides information about modern concepts of PT SCWR.
|Maximum cladding temperature||ºC||850||630||650||700|
In terms of the neutron spectrum, most SCWR designs are a thermal spectrum; however, fast-spectrum SCWR designs are possible (Oka et al., 2010). In general, various liquid or solid moderator options can be utilized in thermal-spectrum SCWRs. These options include light-water, heavy-water, graphite, beryllium oxide, and zirconium hydride. The liquid-moderator concept can be used in both PV and PT SCWRs. The only difference is that in a PV SCWR, the moderator and coolant are the same fluid. Thus, light-water is a practical choice for the moderator. In contrast, in PT SCWRs the moderator and coolant are separated. As a result, there are a variety of options in PT SCWRs.
One of these options is to use a liquid moderator such as heavy-water. One of the advantages of using a liquid moderator in PT SCWRs is that the moderator acts as a passive heat sink in the event of a Loss Of Coolant Accident (LOCA). A liquid moderator provides an additional safety feature Currently, such option is used in CANDU-6 reactors.
Currently, such option is used in CANDU-6 reactors.
The second option is to use a solid moderator. Currently, in RBMK reactors and some other types of reactors such as Magnox, AGR, and HTR, graphite is used as a moderator. However, graphite may catch fire at high temperatures at some conditions. Therefore, other materials such as beryllium, beryllium oxide and zirconium hydride may be used as solid moderators. In this case, heat losses can be reduced significantly. On the contrary, the solid moderators do not act as a passive-safety feature.
High operating temperatures in SCWRs lead to high fuel centreline temperatures. Currently, UO2 has been used in LWRs, PHWRs, etc. However, the uranium-dioxide fuel has a lower thermal conductivity, which results in high fuel centerline temperatures. Therefore, alternative fuels with high thermal-conductivities such as UO2-BeO, UO2-SiC, UO2 with graphite fibre, UC, UC2, and UN might be used (Peiman et al., 2012).
However, the major problem for SCWRs development is reliability of materials at high pressures and temperatures, high neutron flux and aggressive medium such as supercritical water. Unfortunately, up till now nobody has tested candidate materials at such severe conditions.
Major sources for electrical-energy production in the world are: 1) thermal - primary coal and secondary natural gas; 2) nuclear and 3) hydro.
In general, the major driving force for all advances in thermal and nuclear power plants is thermal efficiency. Ranges of gross thermal efficiencies of modern power plants are as the following: 1) Combined-cycle thermal power plants – up to 62%; 2) Supercritical-pressure coal-fired thermal power plants – up to 55%; 3) Carbon-dioxide-cooled reactor NPPs – up to 42%; 4) Sodium-cooled fast reactor NPP – up to 40%; 5) Subcritical-pressure coal-fired thermal power plants – up to 38%; and 6) Modern water-cooled reactors – 30 – 36%.
In spite of advances in coal-fired thermal power-plants design and operation worldwide they are still considered as not environmental friendly due to producing a lot of carbon-dioxide emissions as a result of combustion process plus ash, slag and even acid rains.
Combined-cycle thermal power plants with natural-gas fuel are considered as relatively clean fossil-fuel-fired plants compared to coal and oil power plants, but still emits a lot of carbon dioxide due to combustion process.
Nuclear power is, in general, a non-renewable source as the fossil fuels, but nuclear resources can be used significantly longer than some fossil fuels plus nuclear power does not emit carbon dioxide into atmosphere. Currently, this source of energy is considered as the most viable one for electrical generation for the next 50 – 100 years.
However, all current and oncoming Generation III+ NPPs are not very competitive with modern thermal power plants in terms of thermal efficiency, the difference in values of thermal efficiencies between thermal and nuclear power plants can be up to 20 – 25%.
Therefore, new generation (Generation IV) NPPs with thermal efficiencies close to those of modern thermal power plants, i.e., within a range of 45 – 50% at least, should be designed and built in the nearest future.
ρ density, kg/m3
s, sat saturation
ABWR Advanced Boiling Water Reactor
AECL Atomic Energy of Canada Limited
AGR Advanced Gas-cooled Reactor
BN Fast Neutrons (reactor) (in Russian abbreviation)
BWR Boiling Water Reactor
CANDU CANada Deuterium Uranium
DOE Department Of Energy (USA)
EGP Power Heterogeneous Loop reactor (in Russian abbreviations)
EU European Union
GCR Gas-Cooled Reactor
GFR Gas Fast Reactor
HP High Pressure
HTR High Temperature Reactor
ID Inside Diameter
IP Intermediate Pressure
KAERI Korea Atomic Energy Research Institute (South Korea)
LFR Lead-cooled Fast Reactor
LGR Light-water Graphite-moderated Reactor
LMFBR Liquid-Metal Fast-Breeder Reactor
LP Low Pressure
LWR Light-Water Reactor
MOX Mixed OXides
NIKIET Research and Development Institute of Power Engineering (in Russian abbreviations) or RDIPE, Moscow, Russia
NIST National Institute of Standards and Technology (USA)
NPP Nuclear Power Plant
NRC National Regulatory Commission (USA)
NRU National Research Universal (reactor), AECL, Canada
PCh Pressure Channel
PHWR Pressurized Heavy-Water Reactor
PT Pressure Tube
PV Pressure Vessel
PWR Pressurized Water Reactor
RBMK Reactor of Large Capacity Channel type (in Russian abbreviations)
RPV Reactor Pressure Vessel
SCW SuperCritical Water
SCWR SuperCritical Water Reactor
SFR Sodium Fast Reactor
UK United Kingdom
USA United States of America
VHTR Very High Temperature Reactor
VVER Water-Water Power Reactor (in Russian abbreviation)
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- See some explanations on supercritical-pressures specifics at the end of this section.
- In this reactor the fuel-rod sheath is made of magnesium alloy known by the trade name as “Magnox”, which was used as the name of the reactor (Hewitt and Collier, 2000).
- After the Chernobyl NPP severe nuclear accident in Ukraine in 1986 with the RBMK reactor, graphite is no longer considered as a possible moderator in any water-cooled reactors.
- For the reheat the primary steam is used. Therefore, the reheat temperature is lower than the primary steam temperature. In general, the reheat parameters at NPPs are significantly lower than those at thermal power plants.
- Currently, such option is used in CANDU-6 reactors.