A novel combined small modular reactor (SMR) and gas turbine cycle is presented. This SMR-GT cycle is modeled using fundamental thermodynamic relationships and compared to existing state-of-the-art power generation cycles. The SMR-GT cycle includes an 82 MWe SMR cycle that is combined with a 54 MWe gas turbine cycle. A heat exchanger is used to extract energy from the gas turbine exhaust to create superheated main steam and provide reheat downstream of the LP turbine. This results in a 32 MWe increase in the SMR cycle for total unit output of 136 MWe. Comparisons of thermal efficiency, heat rate, CO2 emissions, and net generation are made between a stand-alone SMR, a typical combined cycle gas turbine (CCGT), standalone gas turbine and the combined SMR-GT cycles. Several advantages of the SMR-GT cycle are discussed. In addition, the rapid deployment of a gas turbine allows for a power station to deliver power and earn revenue prior to completion of the more complex SMR portion of the plant. The SMR portion of the cycle also reduces the overall fuel cost volatility associated with gas turbine based power station.

A new conceptual design of intermediate heat exchanger (IHX) is proposed for application to the gas turbine high temperature reactor system (GTHTR300C) which is being developed by Japan Atomic Energy Agency (JAEA). The GTHTR300C cogenerates hydrogen using the iodine-sulfur (IS) hydrogen production process and electric power using gas turbine. The IHX is used to transport high temperature heat from the nuclear reactor to the hydrogen plant. The IHX proposed in this paper is a horizontal design as opposed to conventional vertical design. Therefore, JAEA investigated the advantage of the horizontal IHX and the economic evaluation when scaling up from conceptual design of high temperature engineering test reactor (HTTR) / IHX to GTHTR300C. To meet the performance requirement, both thermal and structural designs were performed to select heat transfer tube length, tube bundle diameter, insulation thickness, and the length of shell support in a horizontal pressure vessel. It is found that the length of the heat exchanger tube can be shortened and the superalloy-made center pipe structure can be eliminated, which results in reducing the quantity of construction steel by about 30%. Furthermore, the maximum stress concentration in the tubes is found to be significantly reduced such that the creep strength to withstand continuous operation is extended to 40 years, equaling the nuclear reactor life time, without replacement.

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO 2 due to the behavior CO 2 below its critical conditions.

Studies are currently on-going on the cycle performance of Generation IV (Gen IV) Nuclear Power Plants (NPPs) for the purpose of determining optimum operating conditions for efficiency and economic reasons. For Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs), the cycle layout is predominantly driven by the choice of components, the component configuration and the coolant. The purpose of this paper to present and review the cycles currently being considered — the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). In all cases, the cycles utilise helium as the coolant in a closed Brayton gas turbine configuration. Comparisons between the cycles are made for Design Point (DP) and Off-Design Point (ODP) analyses to emphasise the benefits and drawbacks of each cycle. The paper also talks about future trends which include higher Core Outlet Temperatures in excess of 1000 degrees Celsius and the proposal of a simplified cycle configuration which eliminates the need for the recuperator.

A framework – NuTERA (Nuclear Techno-Economic and Risk Assessment) has been developed to set out the requirements for evaluating Generation IV (Gen IV) Nuclear Power Plants (NPPs) at the design conceptual stage. The purpose of the framework is to provide guidelines for future tools that are required to support the decision-making process on the choice of Gen IV concepts and cycle configurations. In this paper, the underpinning of the framework has been demonstrated to enable the creation of an analyses tool, which evaluates the design of an NPP that utilises helium closed Brayton gas turbine cycles. The tool at the broad spectrum focuses on the component and cycle design, Design Point (DP) and Off-Design Point (ODP) performance, part power and load following operations. Specifically, the design model has been created to provide functionalities that look at the in-depth sensitivities of the design factors and operation that affect the efficiency of an NPP such as temperature and pressure ratios, inlet cycle temperatures, component efficiencies, pressure losses. The ODP performance capabilities include newly derived component maps for the reactor, intercooler and recuperator for long term Off-Design (OD) operation. With regard to short term OD, which is typically driven by changes in ambient conditions, the ability to analyse the cycle load following capabilities are possible. An economic model has also been created, which calculates the component costs and the baseline economic evaluation. An incorporated risk model quantifies the performance, operational, financial and design impact risks. However, the tool is able to optimise the NPP cycle configuration based on the best economics using the Levelised Unit Electricity Cost (LUEC) as a measure. The tool has been used to demonstrate a typical decision-making process on 2 Gen IV helium closed gas turbine cycles, which apply to the Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs). The cycles are the Simple Cycle Recuperator (SCR) and Intercooled Cycle Recuperator (ICR). The tool was able to derive the most efficient cycle configurations for the ICR (53% cycle efficiency) and SCR (50% cycle efficiency). Based on these efficiency figures, the baseline LUEC ($/MWh) for the year 2020 is $62.13 for the ICR and $61.84 for the SCR. However, the inclusion of the cost of contingencies due to risks and the subsequent economic optimisation resulted in a cost of $69.70 and $69.80 for the ICR and SCR respectively.

A Very High Temperature Reactor (VHTR) is one of the next generation nuclear reactor systems. From a view point of safety characteristics, a passive cooling system should be designed as the best way of a reactor vessel cooling system (VCS) in the VHTR. Therefore, the gas cooling system with natural circulation is considered as a candidate for the VCS of the VHTR. Japan Atomic Energy Agency (JAEA) is advancing the technology development of the VHTR and is now pursuing design and development of commercial systems such as the 300MWe gas turbine high temperature reactor GTHTR300C (Gas Turbine High Temperature Reactor 300 for Cogeneration). In the VCS of the GTHTR300C, many rectangular flow channels are formed around the reactor pressure vessel (RPV), and a cooling panel utilizing natural convection of air has been proposed. In order to apply the proposed panel to the VCS of the GTHTR300C, it is necessary to clarify the heat transfer and flow characteristics of the proposed channel in the cooling panel. Thus, we carried out an experiment to investigate heat transfer and fluid flow characteristics by natural convection in a vertical rectangular channel heated on one side. Experiments were also carried out to investigate the heat transfer and fluid flow characteristics by natural convection when a porous material with high porosity is inserted into the channel. An experimental apparatus is a vertical rectangular flow channel with a square cross section in which one surface is heated by a rubber heater. Dimensions of the experimental apparatus is 600 mm in height and 50 mm on one side of the square cross section. Air was used as a working fluid and fine copper wire (diameter: 0.5 mm) was used as a porous material. The temperature of the wall surface and gas in the channel were measured by K type thermocouples. The flow velocity distribution was obtained by a PIV method. In this paper, we discuss the heat transfer and fluid flow characteristics of the proposed channel. From the results obtained in the experiment, it was found that the amount of removed heat decreased with increasing of temperature of gas when a copper wire was inserted into the channel with high porosity. This is because the mass flow rate decreased with increasing of viscosity of gas. Since it is expected that the porosity of a porous material will have an optimum value, further studies will be needed.

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs. Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle. Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N 2 mixture, application of SuperCritical (SC)-CO 2 cycles or combined cycles. The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO 2 is provided.

One major challenge to the accurate development of performance simulation tool for component-based nuclear power plant engine models is the difficulty in accessing component performance maps; hence, researchers or engineers often rely on estimation approach using various scaling techniques. This paper describes a multi-fluid scaling approach used to determine the component characteristics of a closed-cycle gas turbine plant from an existing component map with their design data, which can be applied for different working fluids as may be required in closed-cycle gas turbine operations to adapt data from one component map into a new component map. Each component operation is defined by an appropriate change of state equations which describes its thermodynamic behavior, thus, a consideration of the working fluid properties is of high relevance to the scaling approach. The multi-fluid scaling technique described in this paper was used to develop a computer simulation tool called GT-ACYSS, which can be valuable for analyzing the performance of closed-cycle gas turbine operations with different working fluids. This approach makes it easy to theoretically scale existing map using similar or different working fluids without carrying out a full experimental test or repeating the whole design and development process. The results of selected case studies show a reasonable agreement with available data.

Using supercritical carbon dioxide (SCO 2 ) as the working fluid of a closed Brayton cycle gas turbine is widely recognized nowadays, because of its compact layout and high efficiency for modest turbine inlet temperature. It is an attractive option for geothermal, nuclear and solar energy conversion. Compressor is one of the key components for the supercritical carbon dioxide Brayton cycle. With established or developing small power supercritical carbon dioxide test loop, centrifugal compressor with small mass flow rate is mainly investigated and manufactured in the literature; however, nuclear energy conversion contains more power, and axial compressor is preferred to provide SCO 2 compression with larger mass flow rate which is less studied in the literature. The performance of the axial supercritical carbon dioxide compressor is investigated in the current work. An axial supercritical carbon dioxide compressor with mass flow rate of 1000kg/s is designed. The thermodynamic region of the carbon dioxide is slightly above the vapor-liquid critical point with inlet total temperature 310K and total pressure 9MPa. Numerical simulation is then conducted to assess this axial compressor with look-up table adopted to handle the nonlinear variation property of supercritical carbon dioxide near the critical point. The results show that the performance of the design point of the designed axial compressor matches the primary target. Small corner separation occurs near the hub, and the flow motion of the tip leakage fluid is similar with the well-studied air compressor. Violent property variation near the critical point creates troubles for convergence near the stall condition, and the stall mechanism predictions are more difficult for the axial supercritical carbon dioxide compressor.

Closed Brayton cycle (CBC) coupled with High Temperature Gas-cooled Reactor (HTGR) has potential application due to its compact configuration, high power generation efficiency and inherent safety. It is also one of the major power conversion methods for Generation IV advanced nuclear power systems. The typical CBC has several helium-water heat exchangers, including pre-cooler and inter-cooler. These helium-water heat exchangers have important influence on the performance of power conversion system, especially in loss-of-flow accidents (LOFAs). A system model including the reactor and the energy conversion system was established in this paper. The 10MW High Temperature Gas-cooled reactor-test Module helium Gas Turbine (HTR-10GT) was taken as the example to show the consequences of LOFAs. The results showed that LOFAs led to the rising of water temperature out of heat exchangers. With the reduction of water flow rate, the maximum water temperature would increase sharply, and the water temperature in pre-cooler was higher than that in inter-cooler. At low water flow rate, the water temperature would exceed the boiling point. LOFAs also made the rising of helium temperature. It had impacts on the performance of helium compressors. The elevated inlet temperature of helium compressors changed the corrected speed and corrected flow rate, therefore caused the deterioration of compressor’s performance. Furthermore, the LOFAs caused the reactor inlet temperature increasing. In low water flow rate, it would make the reactor inlet temperature beyond the temperature limitation of reactor pressure vessel and influence the safety of reactor. And the LOFAs also reduced the output work of cycle. This paper provides insights of features of CBC in LOFAs and will be helpful to the design and safety operation of closed Brayton cycle coupled with HTGR.

In this paper, the emergency power supply system for next generation sodium fast reactor (SFR) is investigated to target the latest system design with such as optimized equipment for measure of loss of heat removal system (LOHRS) events, taking into account for strengthening diversity and multiplicity. In addition to diversity and multiplicity, optimization in terms of amount of equipment is also taken into account. The emergency power supply system is consist of emergency generators for design based accident (DBA) and emergency generators to cope with the design extension condition (DEC) in order to power the additional cooling equipment of fuel handling system and reactor core cooling system as a measure for LOHRS event. Base on the load capacity and system operation, the configuration of emergency power supply system was examined aiming optimization of amount of equipment in accordance with some shared use taking into account the reliability. As a result, the emergency power supply system with 4 fixed-type gas turbine generators, 4 portable-type diesel generators and batteries was established.

Impinging jet cooling technique has been widely used extensively in various industrial processes, namely, cooling and drying of films and papers, processing of metals and glasses, cooling of gas turbine blades and most recently cooling of various components of electronic devices. Due to high heat removal rate the jet impingement cooling of the hot surfaces is being used in nuclear industries. During the loss of coolant accidents (LOCA) in nuclear power plant, an emergency core cooling system (ECCS) cool the cluster of clad tubes using consisting of fuel rods. Controlled cooling, as an important procedure of thermal-mechanical control processing technology, is helpful to improve the microstructure and mechanical properties of steel. In industries for heat transfer efficiency and homogeneous cooling performance which usually requires a jet impingement with improved heat transfer capacity and controllability. It provides better cooling in comparison to air. Rapid quenching by water jet, sometimes, may lead to formation of cracks and poor ductility to the quenched surface. Spray and mist jet impingement offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronics industries. Mist jet impingement cooling of downward facing hot surface has not been extensively studied in the literature. The present experimental study analyzes the heat transfer characteristics a 0.15mm thick hot horizontal stainless steel (SS-304) foil using Internal mixing full cone (spray angle 20 deg) mist nozzle from the bottom side. Experiments have been performed for the varied range of water pressure (0.7–4.0 bar) and air pressure (0.4–5.8 bar). The effect of water and air inlet pressures, on the surface heat flux has been examined in this study. The maximum surface heat flux is achieved at stagnation point and is not affected by the change in nozzle to plate distance, Air and Water flow rates.

The Design Point (DP) performance of a Nuclear Power Plant (NPP) is fairly straightforward to establish for a given mass flow rate, turbomachinery compressor Pressure Ratio (PR) and reactor Core Outlet Temperature (COT). The plant components are optimum for that point but this is no longer the case if the plant’s operating conditions are changed for part-load performance. Data from tests or previous operating experiences are useful in determining typical part load performance of components based on characteristic maps. However, when individual components are linked in a plant, the range of operating points for part load performances are severely reduced. The main objective of this study is to derive Off-Design Points (ODPs) for the Simple Cycle Recuperated (SCR) and Intercooled Cycle Recuperated (ICR) when considering a temperature range of −35 to 50°C and COTs between 750 to 1000°C, using a modelling & performance simulation tool designed specifically for this study, which calculates the best operational equilibrium ODPs that are critical to the economics of the NPP. Results show that the recuperator High-Pressure (HP) side and reactor pressure losses alter the actual operating parameters (mass flow rate and compressor PR). The SCR yielded a drop in plant cycle efficiency of 1% for a 4% pressure loss in comparison to the ICR (5%) for the same amount of recuperator HP losses. Other parameters such as the precooler and recuperator Low-Pressure (LP) losses still retain the same operating inlet PRs and mass flow rates regardless of the magnitude of the losses. In the absence of characteristic maps in the public domain, the ODPs have been used to produce characteristic trend maps for first order ODP calculations. The analyses intend to aid the development of cycles for Generation IV NPPs specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.

The Intercooled Cycle (IC) is considered as a viable alternative to the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR), based on recent studies in a Nuclear Power Plant (NPP) configuration, which showed plant efficiencies of above 45%. The main difference in configuration is it does not utilise a recuperator. For part load performance, it is expected that the components of the IC will not operate at optimum conditions as the characteristics change. Thus the ability to demonstrate viable part load performance becomes an important requirement for the IC. The main objective of this study is to derive Off-Design Points (ODPs) from a known Design Point (DP) for a temperature range of −35 to 50°C and COTs between 750 to 1000°C. The ODPs have been calculated using a modelling & performance simulation tool designed specifically for this study and aim to provide a set of points that give operational equilibrium, which is critical to the economics of the plant. Results show that the intercooler alters the actual mass flow rate and compressor pressure ratio but the delta across an analysed range of 1 to 5% pressure loss shows a change of ∼9% in plant cycle efficiency, in comparison to the ICR (6%). Furthermore, the reactor pressure losses for IC has the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic trend maps have also been produced for the intercooler operation and the reactor and are applicable for NPP first order calculations. To that effect, it is also proposed to consider the intercooler pressure loss as a handle for ODP performance calculations. The analyses intend to bring further attention to the IC an alternative to current cycle configurations and to aid the development of cycles for Generation IV Nuclear Power Plants specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.

The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear Air-Brayton Combined Cycle systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by Co-Firing when generation is limited. The rated output of a nuclear plant can be augmented by several hundred per cent by Co-Firing. The incremental hydrogen to electricity efficiency can far exceed that of hydrogen in a stand-alone gas turbine.

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and to use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basis accidents and beyond design basis accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake under the loss of the system pressure. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted. At the same time, the system pressure was supposed to lose by some reason. Thus the reactor power increased and the temperature of the core increased. And the protection system warns with two scram signal: too high of the negative varying rate of the system pressure and too high of the reactor power, which should induce the reactor to scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure continued to increase but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1203.4°C. It was a little bit lower than 1230°C — the fuel temperature limitation of HTR-10 and there is no release of any radioactivity. So the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10.

High temperature gas-cooled reactor combined with gas turbine cycle can make full use of the high outlet helium temperature of reactor. Compared to the Open Brayton Cycle, inventory regulation is a new regulation method for Closed Brayton Cycle, which can control the output power of the cycle through the change of pressure level and mass flow of the system. Taking the 10MW high temperature gas-cooled test reactor coupled with helium-turbine cycle (HTR-10GT) for example, a more practice-oriented resistance model is established in this paper for the analysis of influence of resistance characteristic on key components and closed Brayton cycle. When helium inventory drops from 100% to 30% of rated value, the mass flow rate of helium decreases, so the local pressure loss coefficient keeps constant and the pressure loss coefficient increases. As a result, the efficiency of closed cycle decreased about 15%. The performance characteristics of the turbomachinery and heat exchangers are also discussed in detail as key components in the power conversation unit (PCU).

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) have gross thermal efficiencies ranging from 30% and up to 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, fuel – 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 the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MW el Pressure-Channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic 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 CFD Fluent code has been used for better understanding of 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.

High temperature gas-cooled reactor (HTGR) produces not only electricity but also high temperature heat for a variety of heat applications. In the Middle East countries, large demand exists for cogeneration of pure water and electricity from desalination plant coupling with power station. Desalination with an HTGR direct cycle gas turbine system can efficiently meet this demand because such system can produce pure water from using only the waste heat of power generation via secondary pressurized water loop. The waste heat of up to 248MWt is available for desalination from the reactor system of 600 MWt thermal power. In this paper, heat and mass balance was calculated for a new concept of desalination system which is shown to increase use of waste heat by incrementing the number of thermal loading steps at the heat recovery section. Calculation is performed at steady water production condition to clarify the optimum steps of incremental loading based on engineering considerations. As the result, it was found that heat transfer area of heat recovery section in case of brine heater number 3 was 28% smaller than that of number 2.

One of the current engineering challenges is to design next generation (Generation IV) Nuclear Power Plants (NPPs) with significantly higher thermal efficiencies (43–55%) compared to those of current NPPs to match or at least to be close to the thermal efficiencies reached at fossil-fired power plants (55–62%). The Sodium-cooled Fast Reactor (SFR) is one of the six concepts considered under the Generation IV International Forum (GIF) initiative. The BN-600 reactor is a sodium-cooled fast-breeder reactor built at the Beloyarsk NPP in Russia. This concept is the only one from the Generation IV nuclear-power reactors, which is actually in operation (since 1980’s). At the secondary side, it uses a subcritical-pressure Rankine-steam cycle with heat regeneration. The reactor generates electrical power in the amount of 600 MW el . The reactor core dimensions are 0.75 m (height) by 2.06 m (diameter). The UO 2 fuel enriched to 17–26% is utilized in the core. There are 2 loops (circuits) for sodium flow. For safety reasons, sodium is used both in the primary and the intermediate circuits. Therefore, a sodium-to-sodium heat exchanger is used to transfer heat from the primary loop to the intermediate one. In this work major parameters of the reactor are listed. The actual scheme of the power-conversion heat-transport system is presented; and the results of the calculation of thermal efficiency of this scheme are analyzed. Details of the heat-transport system, including parameters of the sodium-to-sodium heat exchanger and main coolant pump, are presented. In this paper two possibilities for the SFR in terms of the power-conversion cycle are investigated: 1. a subcritical-pressure Rankine-steam cycle through a heat exchanger (current approach in Russian and Japanese power reactors); 2. a supercritical-pressure CO 2 Brayton gas-turbine cycle through a heat exchanger (US approach). With the advent of modern super-alloys, the Rankine-steam cycle has progressed into the supercritical region of the coolant and is generating thermal efficiencies into the mid 50% range. Therefore, the thermal efficiency of a supercritical Rankine-steam cycle is also briefly discussed in this paper. According to GIF, the Brayton gas-turbine cycle is under consideration for future nuclear power reactors. The supercritical-CO 2 cycle is a new approach in the Brayton gas-turbine cycle. Therefore, dependence of the thermal efficiency of this SC CO 2 cycle on inlet parameters of the gas turbine is also investigated.