Search

This paper presents optimization of a supercritical carbon dioxide Brayton cycle for three types of 300-MWth small modular reactors (SMRs); a pressurized water reactor (PWR), a sodium-cooled fast reactor (SFR) and a high-temperature gas-cooled reactor (HTGR). The parameters of the pressure ratio and the flow split fraction were examined for sensitivity analysis and optimization of cycle. The optimized cycle efficiencies of PWR, SFR, and HTGR were 30.6%, 46.38%, and 50.04%, respectively. Key components, i.e. turbomachinery and heat exchangers for the SMRs were designed to develop the optimized cycles. The cycle thermal efficiency was improved by using investigating the effects of the channel shape (zigzag, s-shape, airfoil fin) of the printed circuit heat exchangers (PCHEs) on the pressure drop. The study indicated that using airfoil fin type PCHE may increase the cycle thermal efficiency by about 1.0% in comparison with zigzag type PCHE. The effect of turbomachinery efficiencies on the cycle thermal efficiency were investigated.

Nuclear energy is an important ultra-low carbon electricity source. Engineering demonstration reactors, such as Shippingport for Pressurized Water Reactor technology, have been used to advance nuclear reactor technology to deployment. It may be necessary to develop similar engineering demonstration reactors for other reactor concepts that have not yet been built. This paper reviews the historical role of engineering demonstration reactors and highlights the potential for future engineering demonstrations of Fluoride Salt-Cooled High Temperature Reactors (FHRs) and Molten Chloride Fast Reactors (MCFRs). Future engineering demonstration reactors could significantly accelerate deployment of larger-scale FHRs and MCFRs by the advanced nuclear industry via cost and risk reduction. Experience in the United States shows that sustained investment in a public-private partnership, such as the current Advanced Reactor Demonstration Program, is essential for deployment of advanced reactor technologies that have never been built before. The National Reactor Innovation Center at Idaho National Laboratory and Clinch River Site near Oak Ridge National Laboratory are examples of ideal locations for these demonstrations.

It is known that the Pressurized Water Reactors (PWRs), which are the most common type of nuclear reactor existing today, usually used to provide a base load electricity. In order to be able to compete with other generation types (fossil and renewables), it would be desirable to develop PWRs with flexible load following capabilities to cope with varying electricity demand, especially in deregulated markets. The thermal efficiency of PWRs can be increased by fitting the power plant with conventional fossil fuel superheaters. This hybrid system has been hypothesised to be able to adjust the power output and the cycle efficiency of PWRs. Such mode of operation would also improve the efficiency of converting the fossil fuel heat because it is applied only at the superheater stage. There are several ways to supply the heat to the superheaters, for example, by using the exhaust gas from the gas turbines and using the conventional gas burner. In this paper, the thermodynamic performance of the hybrid system (PWR with superheater) is investigated for large reactor and Small Modular Reactor (SMR) application. The thermal efficiency of the AP1000 can be improved from 30.2% to 45.8% (with CCGT), 35.6% (with gas burner), and 36.6% (gas burner with reheating). The thermal efficiency of the SMR can be improved from 33.4% to nearly 45% (with CCGT), 35.5% (with gas burner), and 37.4% (gas burner with reheating). The analysis results show that it is possible for the hybrid system to operate between 65% and the full power load.

PKL is a worldwide unique integral test facility for simulation of thermal-hydraulic phenomena, which can occur in Pressurized Water Reactor under abnormal or accident conditions. The test rig replicates the entire primary side and the relevant parts of the secondary side of western-type PWR plant in the scale of 1:1 in heights and 1:145 in volume and power. The analyses that have been carried out for the last 40 years have always reflected safety issues addressed in the debate of nuclear community. Among goals of experimental programs conducted at the facility there are demonstration of safety margins, investigation of thermal-hydraulic behavior of a plant under accident conditions, verification of accident management procedures as well as provision of data for thermal-hydraulic codes validation. This paper provides a detailed analysis of experiment dedicated to investigation of heat transfer mechanisms in steam generator in presence of non-condensable gas, steam and liquid. Such conditions can occur in a PWR in case of loss of residual heat removal system under cold shutdown conditions. Probabilistic safety assessments indicated that this accident contributes to much larger core damage frequency than originally anticipated. Therefore, an increasing attention has been paid to this accident scenario in the recent years.

A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

Thermoeconomic optimization of a typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant coupled to a Multi Effect Distillation (MED) desalination system with thermo-vapor compressor (TVC) is performed. A thermodynamic modeling based on the energy and exergy analysis is performed while economic modeling is developed based on the Total Revenue Requirement (TRR) method. The objective function based on the thermoeconomic analysis is obtained. The proposed cogeneration plant, for simultaneous production of power and fresh water, including sixteen decision variables is proposed for thermoeconomic optimization in which the goal is minimizing the cost of system product (including the cost of generated electricity and fresh water). The optimization process is performed using a stochastic/deterministic optimization approach namely as Genetic Algorithm. It is found that thermoeconomic optimization aims at reduction of sub-components total costs by reducing either the cost of inefficiency or the cost of owning the components, whichever is dominant. For some components such as evaporators, the improvement is obtained by reducing the owning cost of the sub-system at the cost of reduction of the thermodynamic efficiency. For components like as TVC + de-superheater, improvement is achieved by increasing the thermodynamic efficiency or decreasing the inefficiency cost.

A general survey is presented of energy supplies in the Arab countries. Technologies to exploit nuclear resources are needed in order to ensure a secure energy future. The advantages of introducing nuclear energy into the Arab countries are discussed. The main features of the CANDU (Canada deuterium-uranium) reactors are presented with an assessment concerning their suitability for energy supplies and developing programs. The main differences between the CANDU and PWR (pressurized water reactor) are highlighted.