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).

Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle

2009 ◽  
Author(s):  
V. G. Razumovskiy ◽  
Eu. N. Pis’mennyy ◽  
A. Eu. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Annular Channel ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter

The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.

Numerical and Experimental Investigations on Heat Transfer Phenomena of an Aluminium Microchannel Heat Sink

2011 ◽  
Vol 145 ◽  
pp. 129-133 ◽  
Author(s):  
Thanhtrung Dang ◽  
Ngoctan Tran ◽  
Jyh Tong Teng
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Mass Flow Rate ◽  
Mass Flow ◽  
Flow Rate ◽  
Heat Sink ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Microchannel Heat Sink ◽  
Maximum Heat

The study was done both numerically and experimentally on the heat transfer behaviors of a microchannel heat sink. The solver of numerical simulations (CFD - ACE+software package) was developed by using the finite volume method. This numerical method was performed to simulate for an overall microchannel heat sink, including the channels, substrate, manifolds of channels as well as the covered top wall. Numerical results associated with such kinds of overall microchannel heat sinks are rarely seen in the literatures. For cases done in this study, a heat flux of 9.6 W/cm2was achieved for the microchannel heat sink having the inlet temperature of 25 °C and mass flow rate of 0.4 g/s with the uniform surface temperature of bottom wall of the substrate of 50 °C; besides, the maximum heat transfer effectiveness of this device reached 94.4%. Moreover, in this study, when the mass flow rate increases, the outlet temperature decreases; however, as the mass flow rate increases, the heat flux of this heat sink increases also. In addition, the results obtained from the numerical analyses were in good agreement with those obtained from the experiments as well as those from the literatures, with the maximum discrepancies of the heat fluxes estimated to be less than 6 %.

Heat Transfer to Supercritical Water in Vertical 7-Rod Bundle

2008 ◽  
Author(s):  
V. G. Razumovskiy ◽  
E. N. Pis’mennyy ◽  
A. E. Koloskov ◽  
I. L. Pioro
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Supercritical Water ◽  
Heat Fluxes ◽  
Inlet Temperature ◽  
Outlet Temperature ◽  
Rod Bundle ◽  
Temperature Regimes ◽  
Outside Diameter ◽  
Heated Length

The results of heat transfer to supercritical water flowing upward in a vertical tight 7-rod bundle consisting of tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 700 to 1500 kg/m2s, inlet temperature from 125 to 325°C, outlet temperature up to 379°C and heat flux up to 1.6 MW/m2 (heat flux rate up to 1.5 kJ/kg). Temperature regimes of the bundle cooled by supercritical water were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of the fuel bundles.

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 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.

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.

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.

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