Seismic Response of High Temperature Gas-Cooled Reactor Core with Block-Typed Fuel, (V)

Several efforts have been considered in the development of the modular High Temperature Gas cooled Reactor (HTGR) planned to be a safe and efficient nuclear energy source for the production of electricity and industrial applications. In this work, the RELAP5-3D thermal hydraulic code was used to simulate the steady state behavior of the 10 MW pebble bed high temperature gas cooled reactor (HTR-10), designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET), in China. The reactor core is cooled by helium gas. In the simulation, results of temperature distribution within the pebble bed, inlet and outlet coolant temperatures, coolant mass flow, and others parameters have been compared with the data available in a benchmark document published by the International Atomic Energy Agency (IAEA) in 2013. This initial study demonstrates that the RELAP5-3D model is capable to reproduce the thermal behavior of the HTR-10.

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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.

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Aseismic study of high temperature gas-cooled reactor core with block-type fuel. 3rd report Effect of excitation direction and side support stiffness.

TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series C ◽

10.1299/kikaic.51.746 ◽

1985 ◽

Vol 51 (464) ◽

pp. 746-755

Author(s):

Takeshi IKUSHIMA ◽

Toshiaki HONMA

Keyword(s):

High Temperature ◽

Reactor Core ◽

Block Type ◽

Type Fuel ◽

High Temperature Gas

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Numerical modelling of modular high-temperature gas-cooled reactors with thorium fuel

Nukleonika ◽

10.2478/nuka-2021-0020 ◽

2021 ◽

Vol 66 (4) ◽

pp. 133-138

Author(s):

Mikołaj Oettingen ◽

Jerzy Cetnar

Keyword(s):

High Temperature ◽

Neutron Transport ◽

Reactor Core ◽

Computation Time ◽

Homogenization Method ◽

Uranium Fuel ◽

Neutron Multiplication ◽

3D Numerical Model ◽

Effective Neutron ◽

High Temperature Gas

Abstract The volumetric homogenization method for the simplified modelling of modular high-temperature gas-cooled reactor core with thorium-uranium fuel is presented in the paper. The method significantly reduces the complexity of the 3D numerical model. Hence, the computation time associated with the time-consuming Monte Carlo modelling of neutron transport is considerably reduced. Example results comprise the time evolutions of the effective neutron multiplication factor and fissionable isotopes (233U, 235U, 239Pu, 241Pu) for a few configurations of the initial reactor core.

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