Seawater Desalination Using a High-Temperature Gas-Cooled Reactor (HTR): Overview of Regulatory and Safety Considerations

2020 ◽  
Author(s):  
Mishari Al-Saud ◽  
Fang Chao
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Water Source ◽  
Seawater Desalination ◽  
Holistic View ◽  
Reactor Technology ◽  
Wide Range ◽  
Safety Considerations ◽  
Operational Issues ◽  
High Temperature Gas

Abstract High-Temperature Gas-Cooled Reactor (HTR) is a promising Gen IV reactor technology that has a wide range of applications. Saudi Arabia expressed interest in using HTR as an energy source for seawater desalination. A pre-feasibility study showed that HTR-Desalination is economically competitive and feasible. Yet, the application of HTR power and process heat in the desalination industry faces some technical, conceptual, and regulatory challenges. These challenges are mainly because the reactor and desalination plant are co-located and share common systems and facilities. Moreover, there is a risk of radioactivity and brine discharge impact, since both plants share the water source and discharge location. All these issues challenge the reliability and safety of both plants. Therefore, it is essential to develop effective regulatory frameworks. The basic regulatory and infrastructural requirement for the HTR is like any other nuclear power plant. This study reflects on the typical operational issues and influence of accidents in both plants and their impact on the other. Concluded with regulatory recommendations with an effort to find common interfaces between the regulatory aspects of the nuclear power and desalination industries, which aim at providing a more holistic view on a more comprehensive regulatory framework for nuclear desalination.

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

2008 ◽  
Vol 130 (3) ◽  
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.

Verification of Alarm Displays for the Nuclear Power Plant With Two Modular High-Temperature Gas-Cooled Reactors

2018 ◽  
Author(s):  
Jia Qianqian ◽  
Guo Chao ◽  
Li Jianghai ◽  
Qu Ronghong
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Modular Design ◽  
Early Stage ◽  
Integrated System ◽  
Alarm System ◽  
Control Room ◽  
High Temperature Gas

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.

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

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.

scholarly journals Management of Radioactive Waste Containing Graphite: Overview of Methods

Energies ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 4638
Author(s):  
Leon Fuks ◽  
Irena Herdzik-Koniecko ◽  
Katarzyna Kiegiel ◽  
Grazyna Zakrzewska-Koltuniewicz
Keyword(s):  
High Temperature ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Information Service ◽  
Nuclear Industry ◽  
Current Status ◽  
Triso Fuel ◽  
Irradiated Graphite ◽  
High Temperature Gas

Since the beginning of the nuclear industry, graphite has been widely used as a moderator and reflector of neutrons in nuclear power reactors. Some reactors are relatively old and have already been shut down. As a result, a large amount of irradiated graphite has been generated. Although several thousand papers in the International Nuclear Information Service (INIS) database have discussed the management of radioactive waste containing graphite, knowledge of this problem is not common. The aim of the paper is to present the current status of the methods used in different countries to manage graphite-containing radioactive waste. Attention has been paid to the methods of handling spent TRISO fuel after its discharge from high-temperature gas-cooled reactors (HTGR) reactors.

scholarly journals Using Wireless Sensor Networks to Achieve Intelligent Monitoring for High-Temperature Gas-Cooled Reactor

2017 ◽  
Vol 2017 ◽  
pp. 1-8
Author(s):  
Jianghai Li ◽  
Jia Meng ◽  
Xiaojing Kang ◽  
Zhenhai Long ◽  
Xiaojin Huang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Heterogeneous Data ◽  
Wireless Sensor ◽  
Support Vector ◽  
Intelligent Monitoring ◽  
Novel Algorithms ◽  
Non Gaussian ◽  
High Temperature Gas

High-temperature gas-cooled reactors (HTGR) can incorporate wireless sensor network (WSN) technology to improve safety and economic competitiveness. WSN has great potential in monitoring the equipment and processes within nuclear power plants (NPPs). This technology not only reduces the cost of regular monitoring but also enables intelligent monitoring. In intelligent monitoring, large sets of heterogeneous data collected by the WSN can be used to optimize the operation and maintenance of the HTGR. In this paper, WSN-based intelligent monitoring schemes that are specific for applications of HTGR are proposed. Three major concerns regarding wireless technology in HTGR are addressed: wireless devices interference, cybersecurity of wireless networks, and wireless standards selected for wireless platform. To process nonlinear and non-Gaussian data obtained by WSN for fault diagnosis, novel algorithms combining Kernel Entropy Component Analysis (KECA) and support vector machine (SVM) are developed.

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

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.

scholarly journals Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior

2010 ◽  
Author(s):  
Edwin A. Harvego ◽  
Richard R. Schultz ◽  
Ryan L. Crane
Keyword(s):  
Experimental Data ◽  
High Temperature ◽  
Nuclear Power ◽  
The United States ◽  
Verification And Validation ◽  
Normal Operation ◽  
Nuclear System ◽  
Data Set ◽  
Nuclear Applications ◽  
High Temperature Gas

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes; (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.

Mach stem deformation in pseudo-steady shock wave reflections

2018 ◽  
Vol 861 ◽  
pp. 407-421 ◽  
Author(s):  
Xiaofeng Shi ◽  
Yujian Zhu ◽  
Jiming Yang ◽  
Xisheng Luo
Keyword(s):  
Shock Wave ◽  
High Temperature ◽  
Wall Jet ◽  
Mach Stem ◽  
Wave Reflections ◽  
Reflected Shock ◽  
Wide Range ◽  
Stem Deformation ◽  
Steady Shock ◽  
High Temperature Gas

The deformation of the Mach stem in pseudo-steady shock wave reflections is investigated numerically and theoretically. The numerical simulation provides the typical flow patterns of Mach stem deformation and reveals the differences caused by high-temperature gas effects. The results also show that the wall jet, which causes Mach stem deformation, can be regarded as a branch of the mainstream from the first reflected shock. A new theoretical model for predicting the Mach stem deformation is developed by considering volume conservation. The theoretical predictions agree well with the numerical results in a wide range of test conditions. With this model, the wall-jet velocity and the inflow velocity from the Mach stem are identified as the two dominating factors that convey the influence of high-temperature thermodynamics. The mechanism of high-temperature gas effects on the Mach stem deformation phenomenon are then discussed.

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