The Application of Systems CFD to the Design and Optimization of High-Temperature Gas-Cooled Nuclear Power Plants

ASME 2006 Power Conference ◽

10.1115/power2006-88194 ◽

2006 ◽

Author(s):

Gideon P. Greyvenstein

Keyword(s):

High Temperature ◽

Power Plants ◽

Process Conditions ◽

Great Success ◽

Main Task ◽

Pebble Bed ◽

Design Point ◽

Component Design ◽

Component Models ◽

High Temperature Gas

The basic approach with the design of power plants is to first carry out a thermodynamic cycle analysis and then to vary certain cycle parameters such as the overall pressure ratio in order to determine the optimum or design point condition. One would then proceed to design the different components to match the process conditions. However, since component design has an impact on overall system performance, one cannot optimize the design of the components in isolation from the rest of the system. This calls for an iterative procedure where one has to move several times between the process and component levels to obtain an optimized integrated solution. Another problem faced by plant designers is that Computational Fluid Dynamics (CFD) codes that are increasingly used for detailed component design are slow and not well suited for optimization studies. They are not suited at all for the analysis of complete power plants. Furthermore, the main task of plant designers is not to do design point analyses but to analyze off-design performance, to do uncertainty analyses, to optimize the design and to characterize the dynamic behavior of the system for the purpose of controller design. An approach that has been used with great success for the design of the power conversion system of the Pebble Bed Modular Reactor (PBMR) is the systems CFD approach. The PBMR is a new High Temperature Gas-cooled Reactor (HTR) that is being developed in South Africa. The PBMR utilizes a direct closed recuperated Brayton cycle. Other cycles are also being investigated including various combined cycles. Systems CFD codes are based on the network approach and allow one to model the performance of large complex systems in an integrated fashion. Different levels of component models are provided for ranging from lumped models for components such as pumps to 1D, 2D or even 3D CFD models for components such as complex diffusers, heat exchangers and the pebble bed reactor. In this paper the systems CFD approach will be discussed including the most important component models. Various examples of the application of the systems CFD approach in the design of the PBMR plant will be given.

Download Full-text

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

18th International Conference on Nuclear Engineering: Volume 1 ◽

10.1115/icone18-29055 ◽

2010 ◽

Author(s):

Zhe Dong ◽

Xiaojin Huang ◽

Liangju Zhang

Keyword(s):

Control System ◽

High Temperature ◽

Power Distribution ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Reactor ◽

Reactor Core ◽

Pebble Bed ◽

Nodal Model ◽

High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Download Full-text

The Application of System CFD to the Design and Optimization of High-Temperature Gas-Cooled Nuclear Power Plants

Journal of Engineering for Gas Turbines and Power ◽

10.1115/1.2835057 ◽

2008 ◽

Vol 130 (3) ◽

Cited By ~ 1

Author(s):

Gideon P. Greyvenstein

Keyword(s):

Steady State ◽

High Temperature ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Low Pressure ◽

Critical Parameters ◽

Wide Range ◽

Component Models ◽

High Temperature Gas

The objective of this paper is to model the steady-state and dynamic operation of a pebble-bed-type high temperature gas-cooled reactor power plant using a system computational fluid dynamics (CFD) approach. System CFD codes are 1D network codes with embedded 2D or even 3D discretized component models that provide a good balance between accuracy and speed. In the method presented in this paper, valves, orifices, compressors, and turbines are modeled as lumped or 0D components, whereas pipes and heat exchangers are modeled as 1D discretized components. The reactor is modeled as 2D discretized system. A point kinetics neutronic model will predict the heat release in the reactor. Firstly, the layout of the power conversion system is discussed together with the major plant parameters. This is followed by a high level description of the system CFD approach together with a description of the various component models. The code is used to model the steady-state operation of the system. The results are verified by comparing them with detailed cycle analysis calculations performed with another code. The model is then used to predict the net power delivered to the shaft over a wide range of speeds from zero to full speed. This information is used to specify parameters for a proportional-integral-derivative controller that senses the speed of the power turbine and adjusts the generator power during the startup of the plant. The generator initially acts as a motor that drives the shaft and then changes over to a generator load that approaches the design point value as the speed of the shaft approaches the design speed. A full startup simulation is done to demonstrate the behavior of the plant during startup. This example demonstrates the application of a system CFD code to test control strategies. A load rejection example is considered where the generator load is suddenly dropped to zero from a full load condition. A controller senses the speed of the low pressure compressor/low pressure turbine shaft and then adjusts the opening of a bypass valve to keep the speed of the shaft constant at 60rps. The example demonstrates how detailed information on critical parameters such as turbine and reactor inlet temperatures, maximum fuel temperature, and compressor surge margin can be obtained during operating transients. System CFD codes is a powerful design tool that is indispensable in the design of complex power systems such as gas-cooled nuclear power plants.

Download Full-text

Study on on-line temperature measurement technology for core of pebble bed high temperature gas-cooled reactor

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2020.110944 ◽

2021 ◽

Vol 371 ◽

pp. 110944

Author(s):

Youjie Zhang ◽

Xiang Fang ◽

Tao Ma ◽

Bing Xia ◽

Chuan Li

Keyword(s):

High Temperature ◽

Temperature Measurement ◽

Measurement Technology ◽

Pebble Bed ◽

On Line ◽

Line Temperature ◽

High Temperature Gas

Download Full-text

Benefits, Drawbacks, and Future Trends of Brayton Helium Gas Turbine Cycles for Gas-Cooled Fast Reactor and Very-High Temperature Reactor Generation IV Nuclear Power Plants

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4047773 ◽

2020 ◽

Vol 7 (1) ◽

Author(s):

Arnold Gad-Briggs ◽

Emmanuel Osigwe ◽

Pericles Pilidis ◽

Theoklis Nikolaidis ◽

Suresh Sampath ◽

...

Keyword(s):

High Temperature ◽

Gas Turbine ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Optimum Operating ◽

Future Trends ◽

Design Point ◽

Gen Iv ◽

Very High

Abstract Numerous studies are on-going on to understand the performance of generation IV (Gen IV) nuclear power plants (NPPs). The objective is to determine optimum operating conditions for efficiency and economic reasons in line with the goals of Gen IV. For Gen IV concepts such as the gas-cooled fast reactors (GFRs) and very-high temperature reactors (VHTRs), the choice of cycle configuration is influenced by component choices, the component configuration and the choice of coolant. The purpose of this paper to present and review current cycles being considered—the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). For both cycles, helium is considered as the coolant in a closed Brayton gas turbine configuration. Comparisons are made for design point (DP) and off-design point (ODP) analyses to emphasize the pros and cons of each cycle. This paper also discusses potential future trends, include higher reactor core outlet temperatures (COT) in excess of 1000 °C and the simplified cycle configurations.

Download Full-text

Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

Science and Technology of Nuclear Installations ◽

10.1155/2021/5523104 ◽

2021 ◽

Vol 2021 ◽

pp. 1-10

Author(s):

Jinghan Zhang ◽

Jun Zhao ◽

Jiejuan Tong

Keyword(s):

High Temperature ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Regulatory Commission ◽

Nuclear Safety ◽

Energy Agency ◽

Water Reactor ◽

Regulatory Commission ◽

High Temperature Gas ◽

Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Download Full-text