Influence of the Graphite’s Lifespan on the Design Value of Fuel Burnup in High Temperature Gas-Cooled Reactors

2015 ◽  
Vol 1084 ◽  
pp. 313-316
Author(s):  
Denis F. Baybakov ◽  
Aleksey V. Golovatsky ◽  
Artem G. Naymushin ◽  
Vladimir N. Nesterov ◽  
Savva N. Savanyuk ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Axial Distribution ◽  
Design Value ◽  
Helium Coolant ◽  
High Temperature Gas

This paper describes a method of determining the correlation of the exhausted graphite fuel blocks’ lifespan in high temperature gas-cooled reactors with the fuel burnup. The axial distribution of the local values of the exhausted lifespan of graphite fuel blocks was obtained. It is shown that for ensuring the compliance of the design value of the fuel burnup with graphite fuel blocks operability, it is necessary to reduce the average mixed temperature of the helium coolant leaving the reactor core and as well as reduce the time between nuclear fuel recharges.

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.

Thermal Modeling of the HTR-10 Using the RELAP5-3D Code

2014 ◽  
Author(s):  
Maria Elizabeth Scari ◽  
Antonella Lombardi Costa ◽  
Claubia Pereira ◽  
Clarysson Alberto Mello da Silva ◽  
Maria Auxiliadora Fortini Veloso
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Initial Study ◽  
Industrial Applications ◽  
Pebble Bed ◽  
State Behavior ◽  
Energy Agency ◽  
Helium Gas ◽  
New Energy ◽  
High Temperature Gas

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.

Adsorption removal of carbon dioxide from the helium coolant of high-temperature gas-cooled reactors

1986 ◽  
Vol 60 (4) ◽  
pp. 287-290 ◽  
Author(s):  
A. V. Varezhkin ◽  
Ya. D. Zel'venskii ◽  
I. V. Metlik ◽  
A. A. Khrulev ◽  
A. N. Fedoseenkov
Keyword(s):  
Carbon Dioxide ◽  
High Temperature ◽  
Adsorption Removal ◽  
Helium Coolant ◽  
High Temperature Gas

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

Nukleonika ◽  
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.

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 2019 (3) ◽  
pp. 40-52
Author(s):  
Olga Igorevna Bulakh ◽  
Oleg Konstantinovich Kostylev ◽  
Vladimir Nikolaevich Nesterov ◽  
Eldar Koshalievich Cherdizov
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Fuel Burn ◽  
Design Values ◽  
High Temperature Gas

scholarly journals Development of Analytical Procedures for Chemical Characterization of Substrates for the Production of TRISO Coated Particles as Nuclear Fuel in High Temperature Gas-Cooled Reactors

2020 ◽  
Vol 12 (17) ◽  
pp. 7221
Author(s):  
Ewelina Chajduk ◽  
Paweł Kalbarczyk ◽  
Jakub Dudek ◽  
Marta Pyszynska ◽  
Anna Bojanowska-Czajka ◽  
...  
Keyword(s):  
High Temperature ◽  
Nuclear Fuel ◽  
Inductively Coupled Plasma ◽  
Chemical Characterization ◽  
Quantitative Chemical Analysis ◽  
Coated Particles ◽  
Icp Ms ◽  
Analytical Procedures ◽  
High Temperature Gas

High temperature gas-cooled reactors have recently gained importance as a source of electricity and process heat. Nuclear fuel used in these reactors consists of TRISO (TRiple coated ISOtropic) coated particles, where spherical grains of UO2 or UC2 or UCO kernel are covered with four successive layers consisting of pyrolytic carbon and silicon carbide. Of great importance is the chemical purity of reagents and substances used for the production of TRISO coated fuel particles. Analytical techniques ensuring the determination of elements at trace levels are inductively coupled plasma mass spectrometry (ICP-MS) and neutron activation analysis (NAA). They were applied in this work for the chemical characterization of substrates used for TRISO fuel production. Two analytical procedures were developed: the first, where materials are analyzed using ICP-MS, and the second with the aid of NAA. Successive stages of these procedures are described with details. Results of quantitative chemical analysis of examined substances are reported as well as detection limits for the investigated elements. Moreover, the expanded uncertainties estimated for the determined elements while employing the devised analytical procedures are presented.

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