Goals, Requirements, and Design Implications for the Advanced High-Temperature Reactor

2006 ◽  
Author(s):  
Charles W. Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Power ◽  
Heat Removal ◽  
Security Requirements ◽  
Electricity Production ◽  
Plant Design ◽  
High Temperature Reactor ◽  
High Temperature Gas ◽  
Electrical Output

The Advanced High-Temperature Reactor (AHTR), also called the liquid-salt-cooled Very High-Temperature Reactor (LS-VHTR), is a new reactor concept that has been under development for several years. The AHTR combines four existing technologies to create a new reactor option: graphite-matrix, coated-particle fuels (the same fuel as used in high-temperature gas-cooled reactors); a liquid-fluoride-salt coolant with a boiling point near 1400°C; plant designs and decay-heat-removal safety systems similar to those in sodium-cooled fast reactors; and a helium or nitrogen Brayton power cycle. This paper describes the basis for the selection of goals and requirements, the preliminary goals and requirements, and some of the design implications. For electricity production, the draft AHTR goals include peak coolant temperatures between 700 and 800°C and a maximum power output of about 4000 MW(t), for an electrical output of ∼2000 MW(e). The electrical output matches that expected for a large advanced light-water reactor (ALWR) built in 2025. Plant capital cost per kilowatt electric is to be at least one-third less than those for ALWRs with the long-term potential to significantly exceed this goal. For hydrogen production, the peak temperatures may be as high as 950°C, with a power output of 2400 MW(t). The safety goals are to equal or surpass those of the modular high-temperature gas-cooled reactor with a beyond-design-basis accident capability to withstand large system and structural failures (vessel failure, etc.) without significant fuel failure or off-site radionuclide releases. These safety goals may eliminate the technical need for evacuation zones and reduce security requirements and significantly exceed the safety goals of ALWRs. The plant design should enable economic dry cooling to make possible wider nuclear-power-plant siting options. Uranium consumption is to be less than that for a LWR, with major improvements in repository performance and nonproliferation characteristics.

An Air-Brayton Nuclear-Hydrogen Combined-Cycle Peak- and Base-Load Electric Plant

2007 ◽  
Author(s):  
Charles Forsberg
Keyword(s):  
High Temperature ◽  
Power Output ◽  
Nuclear Reactor ◽  
Peak Load ◽  
Combined Cycle ◽  
Operating Conditions ◽  
Electricity Production ◽  
Spinning Reserve ◽  
High Temperature Reactor ◽  
Base Load

A combined-cycle power plant is proposed that uses heat from a high-temperature nuclear reactor and hydrogen produced by the high-temperature reactor to meet base-load and peak-load electrical demands. For base-load electricity production, air is compressed; flows through a heat exchanger, where it is heated to between 700 and 900°C; and exits through a high-temperature gas turbine to produce electricity. The heat, via an intermediate heat-transport loop, is provided by a high-temperature reactor. The hot exhaust from the Brayton-cycle turbine is then fed to a heat recovery steam generator that provides steam to a steam turbine for added electrical power production. To meet peak electricity demand, after nuclear heating of the compressed air, hydrogen is injected into the combustion chamber, combusts, and heats the air to 1300°C—the operating conditions for a standard natural-gas-fired combined-cycle plant. This process increases the plant efficiency and power output. Hydrogen is produced at night by electrolysis or other methods using energy from the nuclear reactor and is stored until needed. Therefore, the electricity output to the electric grid can vary from zero (i.e., when hydrogen is being produced) to the maximum peak power while the nuclear reactor operates at constant load. Because nuclear heat raises air temperatures above the auto-ignition temperatures of the hydrogen and powers the air compressor, the power output can be varied rapidly (compared with the capabilities of fossil-fired turbines) to meet spinning reserve requirements and stabilize the grid.

Environmentally-Friendly HTGR: MHR-50/100—Concept and Characteristics

2011 ◽  
Author(s):  
Isao Minatsuki ◽  
Tomomi Otani ◽  
Katsusuke Shimizu ◽  
Tetsuo Saguchi ◽  
Sunao Oyama ◽  
...  
Keyword(s):  
High Temperature ◽  
Conceptual Design ◽  
Financial Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Power ◽  
Primary System ◽  
Plant Design ◽  
Plant Dynamics ◽  
High Temperature Gas

A business plan and a new concept of the Mitsubishi small-sized High temperature gas-cooled modular Reactors (MHR-50/100) had been developed as reported in a paper at the HTR-2010 conference in Prague. The present paper reports the results of ensuing conceptual design study including updated market researches, improved safety features of the plant, and the plant dynamics analysis. Market researches on Japan, the USA, Southeast Asia and the Middle East have been updated applying the latest energy outlook data. The result shows that the potential market share for the type of HTGR (high temperature gas reactor) reactors is expected to be 10–20% in new construction of heat source plants in those market areas. A financial analysis made based on the results of the updated market research and the plant cost evaluations indicates that the feasibility of an HTGR business potentially exists. Concerning about the conceptual design, as main themes of the study, a plant design, safety design and plant dynamics have been carried out. The MHR-50/100 high safety characteristics have been confirmed based on the results of the following studies as reported in the present paper: (1) An investigation of a safety scenario during occurrence of a Total Black Out event; (2) An analysis of the reactor decay heat removal via a natural circulation. Lastly, the control methods for the reactor and associated steam cycle system for the MHR-50 have been studied. The results show that the reactor power changes can be effectively achieved by controlling the primary system helium flow rate. The ASURA code developed by MHI is used for simulation of such typical plant transients as 10% step load reduction and full load rejection. The results confirm the easy operability and controllability of the plant.

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 Numerical discretization analysis of a HTR steam generator model for the thermal-hydraulics code trace

2014 ◽  
Vol 29 (suppl.) ◽  
pp. 31-38 ◽  
Author(s):  
Markus Esch ◽  
Dietrich Knoche ◽  
Antonio Hurtado
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Electricity Production ◽  
Thermal Hydraulics ◽  
The Past ◽  
Crucial Component ◽  
High Temperature Reactor ◽  
Different Diameters ◽  
Typical Design ◽  
Steam Generator Model

For future high temperature reactor projects, e. g., for electricity production or nuclear process heat applications, the steam generator is a crucial component. A typical design is a helical coil steam generator consisting of several tubes connected in parallel forming cylinders of different diameters. This type of steam generator was a significant component used at the thorium high temperature reactor. In the work presented the temperature profile is being analyzed by the nodal thermal hydraulics code TRACE for the thorium high temperature reactor steam generator. The influence of the nodalization is being investigated within the scope of this study and compared to experimental results from the past. The results of the standard TRACE code are compared to results using a modified Nusselt number for the primary side. The implemented heat transfer correlation was developed within the past German HTR program. This study shows that both TRACE versions are stable and provides a discussion of the nodalization requirements.

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.

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.

Assessment of Control Room Radiological Habitability of High-Temperature Reactor Pebble-Bed Module in Shidao Bay Multi-Reactor Nuclear Power Site

2018 ◽  
Author(s):  
Xinpeng Li ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Source Term ◽  
Meteorological Data ◽  
Dose Calculation ◽  
Dose Assessment ◽  
Specific Information ◽  
Control Room ◽  
Pebble Bed ◽  
High Temperature Reactor

The control room radiological habitability (CRRH) is important for staff safety in a nuclear power plant, which is also a licensing requirement of the High-temperature Reactor Pebble-bed Module (HTR-PM) in China. Meanwhile, the complexity of the dose assessment increases for the multi-reactor site, which put forward higher requirements for building layout. The CRRH is investigated comprehensively for the multi-reactor site at Shidao Bay in this study. For a large-break loss of coolant accident of HTR-PM and CAP1000 in Shidao Bay nuclear power site, this study estimates doses of body, thyroid and skin due to three exposure pathways using NRC-recommended ARCON96 and dose calculation method in RG 1.195. To perform a realistic evaluation, the latest design and site-specific information are utilized as the input parameters, including the unique accidental source term of HTR-PM and the RG1.183-recommended source term of CAP1000, the release and ventilation parameters, the final layout and the meteorological data in a whole year. The evaluation results demonstrate that the individual dose level of staff in the control room is far below the requirement of the regulatory guide, which guarantees the CRRH of HTR-PM. The contribution of primary radionuclides suggests that tellurium and iodine are primary contributors of the inhalation dose of body and thyroid, which is worthy of paying particular attention to the CRRH design in HTR-PM.

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