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-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 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 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 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.

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.

Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)

2013 ◽  
Author(s):  
W. Peiman ◽  
I. Pioro ◽  
K. Gabriel
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Power Distribution ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Outlet Temperature ◽  
Fuel Bundle ◽  
Flux Profile ◽  
Pressure Channel ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactor (SCWR) is one of the six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits. The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.

Steam-Reheat Option for SCWRs

2009 ◽  
Author(s):  
Eugene Saltanov ◽  
Romson Monichan ◽  
Elina Tchernyavskaya ◽  
Igor Pioro
Keyword(s):  
Thermal Efficiency ◽  
Power Plants ◽  
Nuclear Reactors ◽  
Low Cost ◽  
Thermal Power ◽  
Reaction Rates ◽  
Pressure Tube ◽  
Outlet Temperature ◽  
Chemical Production ◽  
Production Of Hydrogen

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30 – 35% to about 45 – 48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency a nuclear steam reheat has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure in experimental reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear-reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the nuclear steam reheat are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

Thermal Design Options of New Pressure Channel for SCWRs

2009 ◽  
Author(s):  
Wargha Peiman ◽  
Kamiel Gabriel ◽  
Igor Pioro
Keyword(s):  
Heat Loss ◽  
Heat Losses ◽  
Thermal Design ◽  
Operating Pressure ◽  
Channel Design ◽  
Fuel Channel ◽  
Candu Reactors ◽  
Lower Temperature ◽  
Design Options ◽  
Pressure Channel

This paper focuses on thermal-design options of a new pressure channel for SuperCritical Water-cooled Reactors (SCWRs). The objectives of this paper are to estimate heat losses from the coolant to the moderator for a preliminary fuel-channel design and to investigate effects of the insulator thickness and moderator pressure on the overall heat losses. In order to fulfill the objectives, the heat losses for an existing reactor were calculated and compared with available values from open literature. These calculations became the basis for calculation of the heat loss for the chosen new fuel-channel design. MATLAB, and NIST REFPROP software were utilized for programming and calculation of thermo-physical properties as needed, respectively. Heat losses for different thicknesses of the ceramic insulator were calculated. These calculations showed that the heat losses for the optimum thickness of insulator, which was calculated to be 7 mm, were about 31 MW. In current CANDU reactors the operating pressure of the moderator is close to the atmospheric pressure; higher operating pressures will allow operation of the moderator at higher temperature while preventing occurrence of boiling in the calandria vessel. Higher moderator temperatures will results in a lower temperature difference between the coolant and the moderator, hence reducing the heat sink from the coolant to the moderator. The effect of the moderator pressure on the heat loss was investigated, which showed that the heat loss can be reduced by increasing the operating pressure of the moderator by approximately 1 MW per 0.1 MPa increase in pressure.

Steam-Reheat Options for Pressure-Tube SCWRs

2010 ◽  
Author(s):  
Eugene Saltanov ◽  
Wargha Peiman ◽  
Amjad Farah ◽  
Krysten King ◽  
Maria Naidin ◽  
...  
Keyword(s):  
Thermal Efficiency ◽  
Power Plants ◽  
Nuclear Reactors ◽  
Thermal Power ◽  
Reaction Rates ◽  
Pressure Tube ◽  
Outlet Temperature ◽  
Chemical Production ◽  
Production Of Hydrogen ◽  
The Usa

Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 40-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30–35% to about 45–48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs. SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. To achieve higher thermal efficiency Nuclear Steam Reheat (NSR) has to be introduced inside a reactor. Currently, all supercritical turbines at thermal power plants have a steam-reheat option. In the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat at subcritical-pressure experimental boiling reactors. There are some papers, mainly published in the open Russian literature, devoted to this important experience. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. It is obvious that implementation of the nuclear steam reheat at subcritical pressures in pressure-tube reactors is easier task than that in pressure-vessel reactors. Some design features related to the NSR are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors with the nuclear steam reheat is feasible and significant benefits can be expected over other thermal-energy systems.

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