**1. Introduction**

122 Nuclear Reactors

Zeitoun, O., and Shoukri, M., 1996 Bubble behavior and mean diameter in subcooled flow

The demand for clean, non-fossil based electricity is growing; therefore, the world needs to develop new nuclear reactors with higher thermal efficiency in order to increase electricity generation and decrease the detrimental effects on the environment. The current fleet of nuclear power plants is classified as Generation III or less. However, these models are not as energy efficient as they should be because the operating temperatures are relatively low. Currently, a group of countries have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The ultimate goal of developing such reactors is to increase the thermal efficiency from what currently is in the range of 30 - 35% to 45 - 50%. This increase in thermal efficiency would result in a higher production of electricity compared to current Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) technologies.

The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts. These concepts are Gas-cooled Fast Reactor (GFR), Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and SuperCritical Water-cooled Reactor (SCWR). These nuclear-reactor concepts differ in their design in aspects such as the neutron spectrum, coolant, moderator, and operating temperature and pressure.

A SuperCritical Water-cooled Reactor can be designed as a thermal-neutron-spectrum or fast-neutron-spectrum system. SCWR operates above the critical point of water which is at a temperature of 374°C and a pressure of 22.1 MPa. The operating pressure of SCWR is 25 MPa and the outlet temperature of the coolant is 550 - 625°C depending on the design chosen by the respective country that is developing it. The primary choice of fuel for SCWR is an oxide fuel while a metallic fuel has been considered as the secondary choice for the fast-neutron-spectrum SCWRs. A supercritical-water Rankine cycle has been chosen as the power cycle (US DOE, 2002). The thermal efficiency of SCWR is in the range of 45 – 50 %. Figure 1 shows a schematic diagram of a SCWR.

Some of the advantages of SCW Nuclear Power Plants (NPPs) over the conventional NPPs include higher thermal efficiency within a range of 45−50% (Pioro and Duffey, 2007) compared to 30 – 35% for the current NPPs, lower capital costs per kWh of electricity, and the possibility

Thermal Aspects of Conventional and Alternative Fuels

mostly due to the separation of the coolant and the moderator.

to the liquid moderator, especially at SCWR conditions.

in SuperCritical Water-Cooled Reactor (SCWR) Applications 125

liquid moderator options can be utilized in thermal-spectrum SCWRs. These options include light-water, heavy-water, graphite, beryllium oxide, and zirconium hydride (Kirillov et al., 2007). This liquid moderator concept can be used in both PV and PT SCWRs. The only difference is that in a PV SCWR, the moderator and the coolant are the same fluid. Thus, light-water is a practical choice for the moderator. In contrast, in PT SCWRs the moderator and the 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 light-water or 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 feature1, which enhances the safety of operation. On the other hand, one disadvantage of liquid moderators is an increased heat loss from the fuel channels

The second option is to use a solid moderator. Currently, in RBMK reactors and some other types of reactors such as AGR and HTR, graphite is used as the moderator. However, graphite may catch fire at high temperatures under some conditions when exposed to water or oxygen. Other materials such as beryllium oxide and zirconium hydride may be used as solid moderators (Kirillov et al., 2007). In this case, heat losses are reduced significantly. On

High operating temperatures of SCWRs leads to high fuel centerline temperatures. Currently, UO2 has been used in Light Water Reactors (LWRs) and Pressurized Heavy Water Reactors (PHWRs); however, it has a low thermal conductivity which may result in high fuel centerline temperatures. Previous studies (Grande et al., 2010; Pioro et al., 2010; Villamere et al., 2009) have shown that the fuel centerline temperatures could exceed the industry limit of 1850°C (Reisch, 2009) when UO2 is used at SCWR conditions. These studies have been conducted based on an average thermal power per channel and have not taken into account the effects of fuel-sheath gap on the sheath and fuel centreline temperatures. Additionally, the possibility of using enhanced thermal-conductivity fuels in SCWRs has not been examined by previous studies. Moreover, previous studies have focused on the fuel without any emphasis on the fuel channel. Therefore, there is a need to investigate the

Heat transfer at supercritical conditions is characterized by changes in the thermophysical properties of the fluid specifically at pseudocritical points. A pseudocritical point exists at a pressure above the critical pressure of a fluid and at a temperature corresponding to the maximum value of the specific heat for this particular pressure (Pioro and Duffey, 2007). The increase in the specific heat reaches its maximum at the critical point and then decreases as the pressure increases. Furthermore, the pseudocritical temperature increases as the pressure increases. For instance, the corresponding pseudocritical temperatures of lightwater at 23 and 25 MPa are approximately 377.5 and 384.9°C, respectively. Nevertheless, as the temperature passes through the pseudocritical temperature, the specific heat increases. This increase in the specific heat of the fluid allows for the deposition of a significant

the contrary, the solid moderators do not provide a passive-safety feature.

potential use of conventional and alternative fuels for future use in SCWRs.

**2. Heat transfer at supercritical conditions** 

1Currently, such option is used in CANDU-6 reactors.

for co-generation of hydrogen. For instance, the copper-chlorine cycle requires steam at temperatures between 500 and 530°C (Naterer et al., 2009, 2010), which is within the operating range of some SCWR designs. These systems work when supercritical water from a reactor flows through a heat exchanger and transfers heat to a low-pressure steam, which becomes a superheated steam. This superheated steam is transferred at the outlet of the heat exchanger to an adjacent hydrogen plant at a lower pressure (Naterer et al., 2009, 2010).

Fig. 1. Schematic diagram of PV SCWR (US DOE, 2002).

In general, SCWRs can be classified based on the neutron spectrum, moderator, or pressure boundary. In terms of the pressure boundary, SCWRs are classified into two categories, a) Pressure Vessel (PV) SCWRs, and b) Pressure Tube (PT) or Pressure Channel (PCh) SCWRs (Oka et al., 2010; Pioro and Duffey, 2007). The PV SCWR requires a thick pressure vessel with a thickness of about 50 cm (Pioro and Duffey, 2007) in order to withstand high pressures. The vast majority of conventional PWRs and BWRs are examples of PV reactors. Figure 1 shows a schematic diagram of a PV SCWR. On the other hand, the core of a PT SCWR consists of distributed pressure channels, with a thickness of 10 - 15 mm, which might be oriented vertically or horizontally, analogous to RBMK and CANDU reactors, respectively. For instance, SCW CANDU (CANada Deuterium Uranium) reactor 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 has not been ruled out; both horizontal and vertical cores are being studied by the Atomic Energy of Canada Limited (AECL) (Diamond, 2010). Nevertheless, PT SCWRs provide a better control of flow and density variations. On the other hand, in PV SCWRs, there is a non-uniform temperature variation of coolant at the outlet of the pressure vessel.

In terms of the neutron spectrum, most SCWR designs are thermal-spectrum; however, fastspectrum SCWR designs are studied. Recently, Liu et al. (2010) have proposed a mixed spectrum SCWR core, which consists of fast and thermal regions. In general, various solid or

for co-generation of hydrogen. For instance, the copper-chlorine cycle requires steam at temperatures between 500 and 530°C (Naterer et al., 2009, 2010), which is within the operating range of some SCWR designs. These systems work when supercritical water from a reactor flows through a heat exchanger and transfers heat to a low-pressure steam, which becomes a superheated steam. This superheated steam is transferred at the outlet of the heat exchanger to

In general, SCWRs can be classified based on the neutron spectrum, moderator, or pressure boundary. In terms of the pressure boundary, SCWRs are classified into two categories, a) Pressure Vessel (PV) SCWRs, and b) Pressure Tube (PT) or Pressure Channel (PCh) SCWRs (Oka et al., 2010; Pioro and Duffey, 2007). The PV SCWR requires a thick pressure vessel with a thickness of about 50 cm (Pioro and Duffey, 2007) in order to withstand high pressures. The vast majority of conventional PWRs and BWRs are examples of PV reactors. Figure 1 shows a schematic diagram of a PV SCWR. On the other hand, the core of a PT SCWR consists of distributed pressure channels, with a thickness of 10 - 15 mm, which might be oriented vertically or horizontally, analogous to RBMK and CANDU reactors, respectively. For instance, SCW CANDU (CANada Deuterium Uranium) reactor 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 has not been ruled out; both horizontal and vertical cores are being studied by the Atomic Energy of Canada Limited (AECL) (Diamond, 2010). Nevertheless, PT SCWRs provide a better control of flow and density variations. On the other hand, in PV SCWRs, there is a

non-uniform temperature variation of coolant at the outlet of the pressure vessel.

In terms of the neutron spectrum, most SCWR designs are thermal-spectrum; however, fastspectrum SCWR designs are studied. Recently, Liu et al. (2010) have proposed a mixed spectrum SCWR core, which consists of fast and thermal regions. In general, various solid or

an adjacent hydrogen plant at a lower pressure (Naterer et al., 2009, 2010).

Fig. 1. Schematic diagram of PV SCWR (US DOE, 2002).

liquid moderator options can be utilized in thermal-spectrum SCWRs. These options include light-water, heavy-water, graphite, beryllium oxide, and zirconium hydride (Kirillov et al., 2007). This liquid moderator concept can be used in both PV and PT SCWRs. The only difference is that in a PV SCWR, the moderator and the coolant are the same fluid. Thus, light-water is a practical choice for the moderator. In contrast, in PT SCWRs the moderator and the coolant are separated. As a result, there are a variety of options in PT SCWRs, mostly due to the separation of the coolant and the moderator.

One of these options is to use a liquid moderator such as light-water or 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 feature1, which enhances the safety of operation. On the other hand, one disadvantage of liquid moderators is an increased heat loss from the fuel channels to the liquid moderator, especially at SCWR conditions.

The second option is to use a solid moderator. Currently, in RBMK reactors and some other types of reactors such as AGR and HTR, graphite is used as the moderator. However, graphite may catch fire at high temperatures under some conditions when exposed to water or oxygen. Other materials such as beryllium oxide and zirconium hydride may be used as solid moderators (Kirillov et al., 2007). In this case, heat losses are reduced significantly. On the contrary, the solid moderators do not provide a passive-safety feature.

High operating temperatures of SCWRs leads to high fuel centerline temperatures. Currently, UO2 has been used in Light Water Reactors (LWRs) and Pressurized Heavy Water Reactors (PHWRs); however, it has a low thermal conductivity which may result in high fuel centerline temperatures. Previous studies (Grande et al., 2010; Pioro et al., 2010; Villamere et al., 2009) have shown that the fuel centerline temperatures could exceed the industry limit of 1850°C (Reisch, 2009) when UO2 is used at SCWR conditions. These studies have been conducted based on an average thermal power per channel and have not taken into account the effects of fuel-sheath gap on the sheath and fuel centreline temperatures. Additionally, the possibility of using enhanced thermal-conductivity fuels in SCWRs has not been examined by previous studies. Moreover, previous studies have focused on the fuel without any emphasis on the fuel channel. Therefore, there is a need to investigate the potential use of conventional and alternative fuels for future use in SCWRs.
