Experimental Demonstration of a Heat Pipe–Stirling Engine Nuclear Reactor

2014 ◽  
Vol 188 (3) ◽  
pp. 229-237 ◽  
Author(s):  
David I. Poston ◽  
Patrick R. McClure ◽  
David D. Dixon ◽  
Marc A. Gibson ◽  
Lee S. Mason
Keyword(s):  
Heat Pipe ◽  
Nuclear Reactor ◽  
Stirling Engine ◽  
Experimental Demonstration

Preliminary feasibility analysis of Heat Pipe Cooled Bimodal Space Nuclear Reactor

2021 ◽  
pp. 103817
Author(s):  
Xiaoyan Tian ◽  
Huaqi Li ◽  
Duoyu Jiang ◽  
Lei Zhu ◽  
Sen Chen ◽  
...  
Keyword(s):  
Heat Pipe ◽  
Nuclear Reactor ◽  
Feasibility Analysis

Numerical analysis of segmented thermoelectric generators applied in the heat pipe cooled nuclear reactor

2021 ◽  
pp. 117949
Author(s):  
Yin ZHANG ◽  
Kailun GUO ◽  
Chenglong WANG ◽  
Simiao Tang ◽  
Dalin ZHANG ◽  
...  
Keyword(s):  
Numerical Analysis ◽  
Heat Pipe ◽  
Nuclear Reactor ◽  
Thermoelectric Generators

Hybrid Dish/Stirling Systems: Combustor and Heat Pipe Receiver Development

2002 ◽  
Vol 124 (2) ◽  
pp. 176-181 ◽  
Author(s):  
Doerte Laing ◽  
Magnus Pa˚lsson
Keyword(s):  
Heat Pipe ◽  
European Commission ◽  
Stirling Engine ◽  
Combustion System ◽  
Hybrid Mode ◽  
Operation Modes ◽  
Engine Design ◽  
Sodium Heat Pipe

A hybrid sodium heat pipe receiver has been developed within the project HYHPIRE, funded 50% by the European Commission. The hybrid receiver was designed for the SBP/LCS 10-kWel dish/Stirling system with the SOLO-161 Stirling engine. Design of the heat pipe receiver and combustion system are described in this paper. The system has been tested successfully in all operation modes. Results and experience from the lab tests in combustion-only mode, as well as results from demonstration testing in the dish in solar-only, gas-only, and hybrid mode on the Plataforma Solar de Almeria (PSA) in Spain, are reported.

Simulation of a High-Temperature Modular Reactor (HTMR) for Power and Coal-to-Liquid Fuel-Cogeneration Plant

2014 ◽  
Author(s):  
Mubenga Carl Tshamala ◽  
Robert T. Dobson
Keyword(s):  
Power Generation ◽  
High Temperature ◽  
Heat Exchanger ◽  
Heat Pipe ◽  
Superheated Steam ◽  
Nuclear Reactor ◽  
Electrical Power ◽  
Cogeneration Plant ◽  
Process Heat ◽  
Pipe Heat

Traditionally nuclear reactor power plants have been optimized for electrical power generation only. In the light of the ever-rising cost of ever-dwindling fossil fuel resources as well the global polluting effects and consequences of their usage, the use of nuclear energy for process heating is becoming increasingly attractive. In this study the use of a so-called cogeneration plant in which a nuclear reactor energy source is simulated using basic equations for the simultaneous production of superheated steam for electrical power generation and process heat, is considered and analyzed. A novel heat pipe heat exchanger is used to generate superheated steam for the process heat which is, in this case, a coal-to-liquid process (CTL). Natural circulation of sodium, via a thermo-syphon, is used in the heat pipe heat exchanger to transfer heat from the hot stream to the cold. The superheated steam for power generation is generated in a separate once-through helical coil steam generator. A 750 °C, 7 MPa helium cooled high-temperature modular reactor (HTMR) has been considered to simultaneously provide steam at 540 °C, 13.5 MPa for the power unit and steam at 430 °C, 4 MPa for a CTL production plant. The simulation and dynamic control of such a cogeneration plant is considered. In particular, a theoretical model of the plant will be simulated with the aim of predicting the transient and dynamic behavior of the HTMR in order to provide guideline for the control of the plant under various operating conditions. It was found that the simulation model captured the behavior of the plant reasonably well and it is recommended that it could be used in the detailed design of plant control strategies. It was also found that using a 1500 MW-thermal HTMR the South African contribution to global pollution can be reduced by 1.58%.

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Mukesh Kumar ◽  
A. K. Nayak ◽  
Sumit V. Prasad ◽  
P. K. Verma ◽  
R. K. Singh ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Low Flow ◽  
Experimental Demonstration ◽  
Integral Test ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Channel Conditions ◽  
Channel Type

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

Sign in / Sign up

Export Citation Format

Share Document