Thermal Aspects of Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

2010 ◽  
Vol 133 (2) ◽  
Author(s):  
Lisa Grande ◽  
Bryan Villamere ◽  
Leyland Allison ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez-Prado ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Alternative Fuels ◽  
Nuclear Reactors ◽  
Operating Parameters ◽  
Inconel 600 ◽  
Uranium Carbide ◽  
Centerline Temperature ◽  
Sheath Material ◽  
Thermal Conductivities

Supercritical water-cooled nuclear reactors (SCWRs) are a Generation IV reactor concept. SCWRs will use a light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). One reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is to increase the thermal efficiency. The thermal efficiency of existing NPPs is between 30% and 35% compared with 45% and 50% of supercritical water (SCW) NPPs. Another benefit of SCWRs is the use of a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing a pressure tube (PT) type SCWR. This concept refers to a 1200-MWel PT-type reactor. Coolant operating parameters are as follows: a pressure of 25 MPa, a channel inlet temperature of 350°C, and an outlet temperature of 625°C. The sheath material and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for the sheath is a zirconium alloy and the fuel is an enriched uranium dioxide (UO2). The sheath-temperature design limit is 850°C, and the industry accepted limit for the fuel centerline temperature is 1850°C. Previous studies have shown that the maximum fuel centerline temperature of a UO2 pellet might exceed this industry accepted limit at SCWR conditions. Therefore, alternative fuels with higher thermal conductivities need to be investigated for SCWR use. Uranium carbide (UC), uranium nitride (UN), and uranium dicarbide (UC2) are excellent fuel choices as they all have higher thermal conductivities compared with conventional nuclear fuels such as UO2, mixed oxides (MOX), and thoria (ThO2). Inconel-600 has been selected as the sheath material due its high corrosion resistance and high yield strength in aggressive supercritical water (SCW) at high-temperatures. This paper presents the thermalhydraulics calculations of a generic PT-type SCWR fuel channel with a 43-element Inconel-600 bundle with UC and UC2 fuels. The bulk-fluid, sheath and fuel centerline temperature profiles, together with a heat transfer coefficient profile, were calculated for a generic PT-type SCWR fuel-bundle string. Fuel bundles with UC and UC2 fuels with various axial heat flux profiles (AHFPs) are acceptable since they do not exceed the sheath-temperature design limit of 850°C, and the industry accepted limit for the fuel centerline temperature of 1850°C. The most desirable case in terms of the lowest fuel centerline temperature is the UC fuel with the upstream-skewed cosine AHFP. In this case, the fuel centerline temperature does not exceed even the sheath-temperature design limit of 850°C.

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

2015 ◽  
Vol 1 (2) ◽  
Author(s):  
W. Peiman ◽  
I. Pioro ◽  
K. Gabriel
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactors ◽  
Outlet Temperature ◽  
Channel Design ◽  
Fuel Channel ◽  
Temperature And Pressure ◽  
Water Reactors ◽  
Pressure Channel ◽  
Centerline Temperature

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.

Thermal Aspects for Uranium Carbide and Uranium Dicarbide Fuels in Supercritical Water-Cooled Nuclear Reactors

2009 ◽  
Author(s):  
Bryan Villamere ◽  
Leyland J. Allison ◽  
Lisa Grande ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez-Prado ◽  
...  
Keyword(s):  
High Temperature ◽  
Yield Strength ◽  
Supercritical Water ◽  
High Yield ◽  
Operating Parameters ◽  
Nuclear Fuels ◽  
High Yield Strength ◽  
Fuel Channel ◽  
Uranium Carbide ◽  
Sheath Material

SuperCritical Water-cooled Reactors (SCWRs) are a Generation IV nuclear reactor concept. Two main SCWR design concepts are Pressure-Vessel (PV) type and Pressure-Tube (PT) type reactors. SCWRs would use light-water coolant at operating parameters set above the critical point of water (22.1 MPa and 374°C). A reason for moving from current Nuclear Power Plant (NPP) designs to SCW NPP designs is that a SCW NPP will have a thermal efficiency of 45 to 50%, a remarkable improvement from the current 30–35%. SCWRs have another added benefits such as a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. can be eliminated. Canada is in the process of conceptualizing an SCW CANDU reactor. This concept refers to a 1200-MWel horizontal pressure-tube type reactor with the following operating parameters: a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Materials and nuclear fuel must be able to withstand these extreme conditions. In general, the primary choice for a fuel is an enriched Uranium Dioxide (UO2). The industry accepted limit for fuel centreline temperature is 1850°C, and previous studies have shown that the fuel centreline temperature of UO2 pellet might exceed this value at certain conditions. Therefore, a thermal conductivity of the fuel must be sufficiently high to transfer large heat flux within a fuel pellet. Also, a sheath material must withstand supercritical pressures and temperatures inside aggressive medium such as supercritical water, so it should be corrosion-resistant, high-temperature and high-yield strength alloy. In general, sheath materials in various SCWR concepts have a temperature design limit up to 850°C. Uranium Carbide and Uranium Dicarbide are excellent fuel choices as they both have higher thermal conductivities compared to conventional nuclear fuels such as uranium oxide, MOX and Thoria. UC and UC2 are high-temperature ceramics. The sheath material being considered is Inconel 600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. To model a generic SCWR fuel channel, a 43-element bundle string was used. In this paper, bulk-fluid, sheath and fuel centreline temperature profiles together with heat transfer coefficient (HTC) profile were calculated along the heated length of a fuel channel. Also, selected thermophysical properties of various nuclear fuels are listed in the present paper.

Thermal Aspects of Using Uranium Nitride in SuperCritical Water-Cooled Nuclear Reactors

2010 ◽  
Author(s):  
Lisa Grande ◽  
Wargha Peiman ◽  
Sally Mikhael ◽  
Bryan Villamere ◽  
Adrianexy Rodriguez-Prado ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Supercritical Water ◽  
Nuclear Reactors ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Bulk Fluid ◽  
Inconel 600 ◽  
Uranium Nitride ◽  
Axial Heat ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactors (SCWRs) utilize a light-water coolant pressurized to 25 MPa with a channel inlet temperature of 350°C and outlet temperature of 625°C. Previous studies have indicated that uranium dioxide (UO2) nuclear fuel may not be suitable for SCWR use, because the maximum fuel centerline temperature might exceed the industry accepted limit of 1850°C. This research paper explores the use of uranium nitride (UN) as an alternative fuel option to UO2 at SuperCritical Water (SCW) conditions. A generic 1200-MWel Pressure-Tube (PT) -type reactor cooled with SCW was used for this thermalhydraulics analysis. The selected fuel option must have a fuel centerline temperature not higher than the industry accepted limit of 1850°C. Furthermore, the sheath (clad) temperature must not exceed the design limit of 850°C. The sheath and bundle geometry were adopted from previous studies. A single fuel channel was modeled using the UN fuel and an Inconel-600 sheath for several Axial Heat Flux Profiles (AHFPs). Uniform, upstream-skewed cosine, cosine and downstream-skewed cosine AHFPs were used. For each AHFP bulk-fluid, sheath and fuel centerline temperatures, and Heat Transfer Coefficient (HTC) profiles were calculated along the heated length of the channel. The calculations show that the UN fuel maintains a centerline temperature well below the industry accepted limit due to its high thermal conductivity at high temperatures. Therefore, the UN nuclear fuel is a viable fuel option for PT-type SCWRs.

Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

2014 ◽  
Author(s):  
W. Peiman ◽  
I. Pioro ◽  
K. Gabriel
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Reactors ◽  
Outlet Temperature ◽  
Channel Design ◽  
Fuel Channel ◽  
Temperature And Pressure ◽  
Water Reactors ◽  
Pressure Channel ◽  
Centerline Temperature

To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%. In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles. A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.

Thermal Aspects of Using Uranium Mononitride Fuel in a SuperCritical Water-Cooled Reactor at Maximum Heat Flux Conditions

2010 ◽  
Author(s):  
Ashley Milner ◽  
Caleb Pascoe ◽  
Hemal Patel ◽  
Wargha Peiman ◽  
Graham Richards ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Thermal Efficiency ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Maximum Heat ◽  
Light Water ◽  
Maximum Heat Flux ◽  
Centerline Temperature ◽  
Uranium Mononitride

Generation IV nuclear reactor technology is increasing in popularity worldwide. One of the six Generation-IV-reactor types are SuperCritical Water-cooled Reactors (SCWRs). The main objective of SCWRs is to increase substantially thermal efficiency of Nuclear Power Plants (NPPs) and thus, to reduce electricity costs. This reactor type is developed from concepts of both Light Water Reactors (LWRs) and supercritical fossil-fired steam generators. The SCWR is similar to a LWR, but operates at a higher pressure and temperature. The coolant used in a SCWR is light water, which has supercritical pressures and temperatures during operation. Typical light water operating parameters for SCWRs are a pressure of 25 MPa, an inlet temperature of 280–350°C, and an outlet temperature up to 625°C. Currently, NPPs have thermal efficiency about of 30–35%, whereas SCW NPPs will operate with thermal efficiencies of 45–50%. Furthermore, since SCWRs have significantly higher water parameters than current water-cooled reactors, they are able to support co-generation of hydrogen. Studies conducted on fuel-channel options for SCWRs have shown that using uranium dioxide (UO2) as a fuel at supercritical-water conditions might be questionable. The industry accepted limit for the fuel centerline temperature is 1850°C and using UO2 would exceed this limit at certain conditions. Because of this problem, there have been other fuel options considered with a higher thermal conductivity. A generic 43-element bundle for an SCWR, using uranium mononitride (UN) as the fuel, is discussed in this paper. The material for the sheath is Inconel-600, because it has a high resistance to corrosion and can adhere to the maximum sheath-temperature design limit of 850°C. For the purpose of this paper, the bundle will be analyzed at its maximum heat flux. This will verify if the fuel centerline temperature does not exceed 1850°C and that the sheath temperature remains below the limit of 850°C.

Thermal Aspects of Using Uranium Dicarbide Fuel in an SCWR at Maximum Heat-Flux Conditions

2010 ◽  
Author(s):  
Caleb Pascoe ◽  
Ashley Milner ◽  
Hemal Patel ◽  
Wargha Peiman ◽  
Graham Richards ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Alternative Fuels ◽  
Elevated Temperatures ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Bulk Fluid ◽  
Maximum Heat ◽  
Fuel Channel ◽  
Sheath Material

There are 6 prospective Generation-IV nuclear reactor conceptual designs. SuperCritical Water-cooled nuclear Reactors (SCWRs) are one of these design options. The reactor coolant in SCWRs will be light water operating at 25 MPa and up to 625°C, actually at conditions above the critical point of water (22.1 MPa and 374°C, respectively). Current Nuclear Power Plants (NPPs) around the world operate at sub-critical pressures and temperatures achieving thermal efficiencies within the range of 30–35%. One of the major advantages of SCWRs is increased thermal efficiency up to 45–50% by utilizing the elevated temperatures and pressures. SuperCritical Water (SCW) behaves as a single-phase fluid. This prevents the occurrence of “dryout” phenomena. Additionally, operating at SCW conditions allows for a direct cycle to be utilized, thus simplifying the steam-flow circuit. The components required for steam generation and drying can be eliminated. Also, SCWRs have the ability to support hydrogen co-generation through thermochemical cycles. There are two main types of SCWR concepts being investigated, Pressure-Vessel (PV) and Pressure-Tube (PT) or Pressure-Channel (PCh) reactors. The current study models a single fuel channel from a 1200-MWel generic PT-type reactor with a pressure of 25 MPa, an inlet temperature of 350°C and an outlet temperature of 625°C. Since, SCWRs are presently in the design phase there are many efforts in determining fuel and sheath combinations suited for SCWRs. The design criterion to determine feasible material combinations is restricted by the following constraints: 1) The industry accepted limit for fuel centreline temperature is 1850°C, and 2) sheath-material-temperature design limit is 850°C. The primary candidate fuel is uranium dioxide. However; previous studies have shown that the fuel centreline temperature of an UO2 pellet might exceed the industry accepted limit for the fuel centreline temperature. Therefore, investigation on alternative fuels with higher thermal conductivities is required to respect the fuel centreline temperature limit. Sheath (clad) materials must be able to withstand the aggressive SCW conditions. Ideal sheath properties are a high-corrosion resistance and high-temperature mechanical strength. Uranium dicarbide (UC2) is selected as a choice fuel, because of its high thermal conductivity compared to that of conventional nuclear fuels such as UO2, Mixed OXide (MOX) and Thoria (ThO2). The chosen sheath material is Inconel-600. This Ni-based alloy has high-yield strength and maintains its integrity beyond the design limit of 850°C. This paper utilizes a generic SCWR fuel channel containing a continuous 43-element bundle string. The bulk-fluid, sheath and fuel-centreline temperature profiles together with Heat Transfer Coefficient (HTC) profile were calculated along the heated length of a fuel channel at the maximum Axial Heat Flux Profiles (AHFPs).

Thermal Aspects of Using Thoria Fuel in SuperCritical Water-Cooled Nuclear Reactors

2009 ◽  
Author(s):  
Lisa Grande ◽  
Bryan Villamere ◽  
Adrianexy Rodriguez-Prado ◽  
Sally Mikhael ◽  
Leyland Allison ◽  
...  
Keyword(s):  
Thermal Conductivity ◽  
Supercritical Water ◽  
Alternative Fuels ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Reactor Design ◽  
Operating Conditions ◽  
Thermal Design ◽  
Suitable Alternative ◽  
Design Options

SuperCritical Water-cooled Reactors (SCWRs) are one of six next-generation nuclear-reactor design options under consideration worldwide. These nuclear-reactor design options are included in the major international treaties such as: Generation IV International Forum (GIF) and INternational PROject on innovative nuclear reactors and fuels (INPRO). SCWR coolant is light water, which operates at supercritical pressures and temperatures. Typical SCWR coolant operating parameters are 25 MPa and 350–625°C. These SCWR operating conditions significantly increase the thermal efficiency of a SCW Nuclear Power Plant (NPP) (about 45 – 50%) compared to that of existing NPPs (30 – 35%). Also, SCWRs use significantly higher water parameters than existing water-cooled reactors, because of this they can support hydrogen co-generation. Previous thermal-design fuel-channel option studies for SCWRs have shown that the use of uranium dioxide (UO2) fuel at supercritical water conditions might be unacceptable as the fuel centerline temperature is close to or even exceeds the industry accepted limit of 1850°C. Alternative fuels with a higher thermal conductivity have to be considered. Thoria (ThO2) fuel is a suitable alternative to UO2 due to its higher thermal conductivity. Thoria fuel is beneficial because it complies with the Non-Proliferation Treaty and there are plenty of reserves worldwide. Therefore, ThO2 fuel and its suitability with SCWR use are considered in this paper.

Thermal Aspects of Using Thorium Dioxide as Alternative Nuclear Fuel in SuperCritical Water-Cooled Reactors

2010 ◽  
Author(s):  
Hemal Patel ◽  
Ashley Milner ◽  
Caleb Pascoe ◽  
Wargha Peiman ◽  
Graham Richards ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Supercritical Water ◽  
Operating Conditions ◽  
Outlet Temperature ◽  
Bulk Fluid ◽  
Thorium Dioxide ◽  
Maximum Heat ◽  
Fuel Channel ◽  
Inconel 600 ◽  
Centerline Temperature

SuperCritical Water-Cooled nuclear Reactors (SCWRs) are one of six choices for Generation IV (Gen IV) reactor concepts. These reactors use light water as a coolant and operate at a pressure of 25 MPa, inlet temperatures 280–350°C and an outlet temperature up to 625°C. Operating at these elevated temperatures and pressures are beneficial due to: 1) increased gross thermal efficiency of SCW Nuclear Power Plants (NPPs) (from 30%–35% of the current NPPs to 45%–50%) and 2) decreased capital and operational costs. Use of SCW as a reactor coolant will permit a direct-cycle steam circuit. SCWRs eliminate the need for steam generators, steam separators, and steam dryers. Another advantage of SCWRs is a possibility for hydrogen co-generation through thermochemical cycles. At these extreme operating conditions we must be ensured that all fuel-channel materials, i.e., sheath (clad) and fuel, will operate below accepted temperature limits. The industry accepted limit for the fuel centerline temperature is 1850°C, and the design limit for sheath temperature is 850°C. Material investigations have begun with existing NPP fuel-channel designs. Previous studies with UO2 fuel at SCW conditions have indicated that the fuel centerline temperature may exceed the temperature limit. Zirconium alloys cannot operate at temperature beyond 350–500°C due to high corrosion rates. Therefore, Inconel-600 was chosen as a sheath material since is maintains a high yield strength and corrosion resistance at high temperatures. Uranium dioxide fuel is widely used and world resources are becoming limited. Thoria or thorium dioxide (ThO2) is considered as an alternative nuclear fuel and offers many benefits. Thorium dioxide is compliant to the Non-Proliferation Treaty, abundant in global reserves and has higher thermal conductivity than that of UO2. An objective of this paper is to determine the suitability of ThO2 fuel in an Inconel-600-sheath fuel bundle within an SCWR fuel channel. Bulk-fluid, outer-sheath and fuel centerline temperature profiles along with Heat Transfer Coefficient (HTC) profiles were computed along the heated length of a bundle string at the maximum heat flux.

Study on Thermal Efficiency of SuperCritical Water NPPs

2016 ◽  
Author(s):  
Cristina Mazza ◽  
Paul Ponomaryov ◽  
Yifeng Zhou ◽  
Igor Pioro
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactors ◽  
Thermal Power ◽  
Vital Role ◽  
Energy Market ◽  
Power Cycle ◽  
Cost Efficient

As the demand for emission-free energy increases, the continued improvement of Nuclear Power Plants (NPPs) and their thermal efficiencies is crucial to fulfilling that demand. Current NPPs, especially, with water-cooled reactors, have significantly lower thermal efficiencies (32–36%) compared to those of modern advanced thermal power plants (55–62%). Even Generation-III+ water-cooled NPPs will have thermal efficiencies not higher than 37–38%. Therefore, to be competitive on the energy market, new nuclear reactors and NPPs, so-called, Generation-IV concepts, should be designed and commissioned. The paper discusses the vital role that thermal efficiency plays with respect to how far nuclear reactors can be more cost efficient and competitive. An evaluation of thermal efficiencies has been carried out for SuperCritical Water (SCW) NPPs with Rankine “steam”-turbine power cycle using the IAEA DEsalination Thermodynamic Optimization Program (DE-TOP). Various options for improving thermal efficiencies of SCW NPPs have been studied. This study was performed in support of possible designs of the first experimental SCW reactors.

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