scholarly journals Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

2019 ◽  
Vol 347 ◽  
pp. 108-121 ◽  
Author(s):  
Cihang Lu ◽  
Nicholas R. Brown
Keyword(s):  
High Temperature ◽  
Fuel Cycle ◽  
Design Basis ◽  
High Temperature Gas

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

Deep Burn Strategy for Pure Reactor-Grade Plutonium in Pebble Bed High Temperature Gas-Cooled Reactors

2008 ◽  
Author(s):  
Eben Mulder ◽  
Dawid Serfontein ◽  
Eberhard Teuchert
Keyword(s):  
High Temperature ◽  
Fuel Cycle ◽  
Waste Incineration ◽  
Β Decay ◽  
Pebble Bed ◽  
Pebble Bed Reactors ◽  
Advanced Fuel ◽  
Very High ◽  
High Temperature Gas

In this article an advanced fuel cycle for pebble bed reactors is introduced that can safely and efficiently incinerate pure reactor-grade Pu [Pu(LWR)], thereby fulfilling the bulk of the GNEP waste incineration requirements. It is shown below that the very high fissile content of the Pu(LWR)-fuel enables it to convert practically all of the 240Pu to 241Pu and incinerate it. Since the fuel contains no 238U, no fresh 239Pu is produced. The 239Pu is reduced in-situ by 99.5% and the 240Pu by 97.6%. The only significant fissile isotope remaining is 241Pu, however, it will decay with a half life of 14.4 years to the fertile 241Am by β-decay.

Thorium Cycle in High Temperature Gas-Cooled Gas Turbine Reactors (HTG-GT) Using Highly Enriched Uranium Obtained From the Dismantling of Nuclear Weapons

1994 ◽  
Author(s):  
Nicola Cerullo ◽  
Giovanni Guglielmini ◽  
A. Di Pietro
Keyword(s):  
High Temperature ◽  
Gas Turbine ◽  
Nuclear Weapons ◽  
Fuel Cycle ◽  
Fissile Material ◽  
High Investment ◽  
Enriched Uranium ◽  
The Cold War ◽  
Thorium Cycle ◽  
High Temperature Gas

The closed thorium fuel cycle is based on the use of fissile U-233 produced by the thorium fertilization in the original fuel element without any refabrication action, which is very difficult, due to the high activity of Thorium activated products. The need of a consistent amount of fissile material for beginning the U-Th cycle activity, in order to sustain the Thorium conversion reactions, requires an high initial U-235 enrichment. This condition, due to high investment costs, stopped, in the last years, any initiative in this field. The end of the cold war and the disarmament agreements pose the problem of the use of military grade fissile materials resulting from the dismantling of nuclear weapons both Russian and American. In this paper the problem is analyzed and a High Temperature Gas-cooled Gas Turbine (HTG-GT) reactor, using a nuclear U-Th fuel cycle utilizing military grade highly enriched uranium, is proposed.

scholarly journals Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

2015 ◽  
Vol 2015 ◽  
pp. 1-8 ◽  
Author(s):  
Igor Shamanin ◽  
Sergey Bedenko ◽  
Yuriy Chertkov ◽  
Ildar Gubaydulin
Keyword(s):  
High Temperature ◽  
3D Model ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Closed Fuel Cycle ◽  
Comparison Results ◽  
Long Time ◽  
Technological Platform ◽  
High Temperature Gas ◽  
Unified Design

Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

scholarly journals Simulation of the HTR-10 Operation History With the PANGU Code

2021 ◽  
Vol 9 ◽  
Author(s):  
Ding She ◽  
Fubing Chen ◽  
Bing Xia ◽  
Lei Shi
Keyword(s):  
High Temperature ◽  
Fuel Cycle ◽  
Outlet Temperature ◽  
Reactor Physics ◽  
Important Data ◽  
Analysis And Design ◽  
Reactor Test ◽  
Test Module ◽  
High Temperature Gas ◽  
Htgr Fuel

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-cooled Reactor (HTGR) in China, which was operated from January 2003 to May 2007. The HTR-10 operation history provides very important data for the validation of HTGR codes. In this paper, the HTR-10 operation history is simulated with the PANGU code, which has been recently developed for HTGR reactor physics analysis and design. Models and parameters are constructed based on the measured data of the actual conditions. The simulation results agree well with the measurements in all steady-state power periods. The discrepancy of keff is generally below 0.5%, and the discrepancy of coolant outlet temperature is generally below 5°C. It is also figured out that the burnup of graphite impurities has considerable influence on the keff at the end of the operation history, which can cause over 1.5% discrepancy when neglecting the burnup of graphite impurities. By this work, the PANGU code’s applicability in actual HTGR fuel cycle simulations is demonstrated.

scholarly journals Extension of lifespan of graphite in fuel blocks of high-temperature gas-cooled reactors as the resource for ensuring design values of nuclear fuel burn-up

2019 ◽  
Vol 5 (4) ◽  
pp. 289-295 ◽  
Author(s):  
Olga I. Bulakh ◽  
Oleg K. Kostylev ◽  
Vladimir N. Nesterov ◽  
Eldar K. Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Alternative Energy ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Calculated Data ◽  
Fuel Burnup ◽  
The Core ◽  
High Temperature Reactors ◽  
High Temperature Gas

High-temperature gas-cooled reactor (HTGR) is one of promising candidates for new generation of nuclear power reactors. This type of nuclear reactor is characterized with the following principal features: highly efficient generation of electricity (thermal efficiency of about 50%); the use of high-temperature heat in different production processes; reactor core self-protection properties; practical exclusion of reactor core meltdown in case of accidents; the possibility of implementation of various nuclear fuel cycle options; reduced radiation and thermal effects on the environment, forecasted acceptability of financial performance with respect to cost of electricity as compared with alternative energy sources. The range of output coolant temperatures in high-temperature reactors within the limits of 750–950 °C predetermines the use of graphite as the structural material of the reactor core and helium as the inert coolant. Application of graphite ensures higher heat capacity of the reactor core and its practical non-meltability. Residence time of reactor graphite depends on the critical value of fluence of damaging neutrons (neutrons with energies above 180 keV). In its turn, the value of critical neutron fluence is determined by the irradiation temperature and flux density of accompanying gamma-radiation. The values of critical fluence for graphite decrease within high-temperature region of 800–1000 °C to 1·1022 – 2·1021 cm–2, respectively. The compactness of the core results in the increase of the fracture of damaging neutrons in the total flux. These circumstances predetermine relatively low values of lifespan of graphite structures in high-temperature reactors. Design features and operational parameters of GT-MHR high-temperature gas-cooled reactor are described in the present paper. Results of neutronics calculations allowing determining the values of damaging neutron flux, nuclear fuel burnup and expired lifespan of graphite of fuel blocks were obtained. The mismatch between positions of the maxima in the dependences of fuel burnup and exhausted lifespan of graphite in fuel blocks along the core height is demonstrated. The map and methodology for re-shuffling fuel blocks of the GT-MHR reactor core were developed as the result of analysis of the calculated data for ensuring the matching between the design value of the fuel burnup and expected total graphite lifespan.

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