Wargha Peiman

Thermal Aspects of Conventional and Alternative Fuels in SuperCritical Water-Cooled Reactor (SCWR) Applications

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.

Thermal Design Options of New Pressure Channel for SCWRs

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.Copyright

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation. The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8 MPa/257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., supercritical water-cooled reactors (SCWRs) have to be designed. This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel pressure-channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.

Sensitivity Analysis of Fuel Centerline Temperature in SCWRs

The Generation IV International Forum (GIF) is intended to encourage the world’s leading nuclear countries to develop nuclear energy systems that can supply future energy demands. There are six nuclear reactor concepts under research and development as part of the GIF. The SuperCritical Water-cooled Reactor (SCWR) is one of these six nuclear-reactor concepts. The proposed SCWRs operate at high temperatures and pressures at around 625°C and 25 MPa, respectively. These high operating parameters are essential in order to achieve a thermal efficiency of around 45–50%, which is significantly higher than those of the current conventional nuclear power plant (NPPs) which operate at a thermal efficiency in the range of 30–35%.The SCWRs high operating temperatures and pressures impose many challenges. One of these challenges is the heating of the fuel to temperatures that can cause fuel melting. The main objective of this paper is to conduct a sensitivity analysis in order to determine the factors mostly affecting the fuel centerline temperature. In this process, different thermal conductivity fuels such as Mixed Oxide Fuel (MOX), Uranium Oxide + Beryllium Oxide (UO2+BeO), and Uranium Carbide (UC) will be examined enclosed in a 54-element fuel bundle. Other factors such as the sheath material and the Heat Transfer Coefficient (HTC) might also affect the fuel centerline temperature. The HTC will be increased by a multiple of two and the fuel centerline temperature will be calculated. Therefore, in this paper the HTC, bulk-fluid, sheath and fuel centerline temperature will be calculated along the heated length of a generic SCWR fuel channel at an average channel thermal power of 8.5 MWth.Copyright

Current Status and Future Applications of Supercritical Pressures in Power Engineering

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. 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. Therefore, thermal and nuclear electrical-energy production as the major source is considered in the paper.From thermodynamics it is well known that higher thermal efficiencies correspond to higher temperatures and pressures. Therefore, modern SuperCritical (SC)-pressure coal-fired power plants have thermal efficiencies within 43–50% and even slightly above. Steam-generator outlet temperatures or steam-turbine inlet temperatures have reached a level of about 625°C (and even higher) at pressures of 25–30 (35–38) MPa. This is the largest application of SC pressures in industry.In spite of advances in coal-fired power-plants 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.The most efficient modern thermal-power plants with thermal efficiencies within a range of 50–60%, are so-called, combined-cycle power plants, which use natural gas as a fuel. Natural gas is considered as a clean fossil fuel compared to coal and oil, but still due to combustion process emits a lot of carbon dioxide when it used for electrical generation. Therefore, a new reliable and environmental friendly source for the electrical-energy generation should be considered.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.Current, i.e., Generation II and III, Nuclear Power Plants (NPPs) consist of water-cooled reactors NPPs with the thermal efficiency of 30–35% (vast majority of reactors); subcritical carbon-dioxide-cooled reactors NPPs with the thermal efficiency up to 42% and liquid-sodium-cooled reactor NPP with the thermal efficiency of 40%. Therefore, the current fleet of NPPs, especially, water-cooled NPPs, are not very competitive compared to modern thermal power plants. Therefore, next generation or Generation-IV reactors with new parameters (NPPs with the thermal efficiency of 43–50% and even higher for all types of reactors) are currently under development worldwide.Generation-IV nuclear-reactor concept such as SuperCritical Water-cooled Reactor (SCWR) is intended to operate with direct or in-direct SC-“steam” Rankine cycle. Lead-cooled Fast Reactor (LFR) can be connected to SC-“steam” Rankine cycle or SC CO2 Brayton cycle through heat exchangers. In general, other Generation IV reactor concepts can be connected to either one or another cycle through heat exchangers.Therefore, this paper discusses various aspects of application of SC fluids in power engineering.Copyright

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

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).Copyright

Thermal aspects of conventional and alternative fuels

The ultimate goal of nuclear safety is to prevent the uncontrolled release of radioactive nuclides from the fuel into the environment. To meet this objective, a number of engineering barriers have been designed, including the fuel matrix and fuel cladding (or sheath). The mechanical integrity of these two barriers should be maintained under normal, transient, and accident conditions. Hence, neither the fuel nor the cladding should reach its melting temperature. To comply with this requirement, safety margins have been established to safeguard the mechanical integrity of the fuel and cladding. In addition, experiments have been conducted, and analytical studies have been undertaken on fuel performance to provide quantifiable measures to ensure operation within the operating limits. Thermophysical, mechanical, and irradiation properties of the nuclear fuels are of paramount importance in these analytical studies. This manuscript provides a literature survey of various nuclear fuels categorized as (1) metallic fuels; (2) ceramic fuels; and (3) composite fuels with a focus on these properties.

Heat-Loss Calculations in a SCWR Fuel-Channel

The objective of this paper is to calculate heat losses from a CANDU-6 fuel-channel while modifying it according to the specified operating pressure and temperature conditions of SuperCritical Water-cooled Reactors (SCWRs). Heat losses from the coolant to the moderator are significant in a SCWR because of high operating temperatures (i.e., 350–625°C). This has adverse effects on the overall thermal efficiency of the Nuclear Power Plant (NPP), so it is necessary to determine the amount of heat losses from fuel-channels proposed for SCWRs. Inconel-718 was chosen as a pressure tube (PT) material and PT minimum required thickness was calculated in accordance with the coolant’s maximum operating pressure and temperature. The heat losses from the fuel-channel were calculated along the heated length of the fuel-channel. Steady-state one-dimensional heat-transfer analysis was conducted, and programming in MATLAB was performed. The fuel-channel was divided into small segments and for each segment thermal resistances of the fuel-channel components were analyzed. Further, the thermophysical properties of the coolant, annulus gas, and moderator were retrieved from the NIST REFPROP software. The analysis outcome resulted in a total heat loss of 29.3 kW per fuel-channel when the pressure of the annulus gas was 0.3 MPa.Copyright

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

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.© 2010 ASME

Steam-Reheat Options for Pressure-Tube SCWRs

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.Copyright